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Sample records for beta tokamak research

  1. [High beta tokamak research

    International Nuclear Information System (INIS)

    Our activities on High Beta Tokamak Research during the past 20 months of the present grant period can be divided into six areas: reconstruction and modeling of high beta equilibria in HBT; measurement and analysis of MHD instabilities observed in HBT; measurements of impurity transport; diagnostic development on HBT; numerical parameterization of the second stability regime; and conceptual design and assembly of HBT-EP. Each of these is described in some detail in the sections of this progress report

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

    International Nuclear Information System (INIS)

    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. Theory of high-beta tokamaks

    International Nuclear Information System (INIS)

    The theoretical researches on high beta tokamak are reviewed. The ballooning mode instability is thought to be the most serious problem for the high beta tokamaks, and the theoretical results on the ballooning mode instability are discussed in detail. The experimental results in high beta belt pinch devices are also discussed. (author)

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

    International Nuclear Information System (INIS)

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

  5. Research using small tokamaks

    International Nuclear Information System (INIS)

    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

  6. Research using small tokamaks

    International Nuclear Information System (INIS)

    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

  7. Research using small tokamaks

    International Nuclear Information System (INIS)

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

  8. Research using small tokamaks

    International Nuclear Information System (INIS)

    This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs

  9. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  10. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  11. Status of tokamak research

    International Nuclear Information System (INIS)

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

  12. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254. ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  13. Observation of finite-. beta. MHD phenomena in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, K.M.

    1984-09-01

    Stable high-beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1 internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of q/sub 0/ and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas, two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded. Whether this represents an ultimate tokamak limit or if beta optimized configurations (Dee- or bean-shaped plasmas) can exceed this limit and perhaps enter a second regime of stability remains to be clarified.

  14. Anomalous diffusion and transport beta limits in dense tokamak plasmas

    International Nuclear Information System (INIS)

    For tokamak plasmas which are sufficiently large and/or dense that the ionization source on axis may be neglected, particle balance is achieved by the inward diffusion due to the Ware pinch compensating the outward flow due to both neoclassical and anomalous diffusion. Insertion of measured data into the particle flux balance relation determines the anomalous particle diffusion coefficient Dsub(A); comparison of the results from a variety of tokamaks implies that the dominant dependence on machine and/or plasma parameters is Dsub(A) proportional to 1/n. Particle flux balance also implies an upper bound on the central value of βsub(e), the limiting value being obtained when the plasma parameters are chosen such that Dsub(A)<< Dsub(NEO). This limit has been computed for circular-cross-section tokamaks, and the results so obtained are of the same order as magnetohydrodynamic beta limits. (author)

  15. TIBER: Tokamak Ignition/Burn Experimental Research. Final design report

    International Nuclear Information System (INIS)

    The Tokamak Ignition/Burn Experimental Research (TIBER) device is the smallest superconductivity tokamak designed to date. In the design plasma shaping is used to achieve a high plasma beta. Neutron shielding is minimized to achieve the desired small device size, but the superconducting magnets must be shielded sufficiently to reduce the neutron heat load and the gamma-ray dose to various components of the device. Specifications of the plasma-shaping coil, the shielding, coaling, requirements, and heating modes are given. 61 refs., 92 figs., 30 tabs

  16. Magnetohydrodynamic instability, feedback stabilization, and disruption study for the Korea superconducting tokamak advanced research tokamak

    International Nuclear Information System (INIS)

    Passive and active feedback stabilization schemes being considered in Korea Superconducting Tokamak Advanced Research (KSTAR) device for the stabilization of the resistive magnetohydrodynamic modes such as the resistive wall and the neoclassical tearing are briefly introduced. A short summary is also presented on the tokamak simulation results of disruption dynamics and load in the KSTAR tokamak obtained using the tokamak simulation code (TSC)

  17. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

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

  18. Beta limits in long-pulse tokamak discharges

    Energy Technology Data Exchange (ETDEWEB)

    Sauter, O.; La Haye, R.J.; Chang, Z.; Gates, D.A.; Kamada, Y.; Zohm, H.; Bondeson, A.; Boucher, D.; Callen, J.D.; Chu, M.S.; Gianakon, T.A.; Gruber, O.; Harvey, R.W.; Hegna, C.C.; Lao, L.L.; Monticello, D.A.; Perkins, F.; Pletzer, A.; Reiman, A.H.; Rosenbluth, M.; Strait, E.J.; Taylor, T.S.; Turnbull, A.D.; Waelbroeck, F.; Wesley, J.C.; Wilson, H.R.; Yoshino, R. [CRPP, Assoc. Euratom-Switzerland, PPB-Ecublens, 1015 Lausanne (Switzerland)]|[ITER-JCT, 11025 N Torrey Pines Rd., La Jolla, California 92037 (United States)]|[General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)]|[Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)]|[UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)]|[JAERI, Naka-machi, Naka-gun, Ibaraki-ken, 311-01 (Japan)]|[MIP fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany)]|[Chalmers University of Technology, Goeteborg (Sweden)]|[University of Wisconsin, Madison, Wisconsin 53706-16787 (United States)]|[CompX, 12839 Via Grimaldi, Del Mar, California 92014 (United States)]|[IFS, University of Texas, Austin, Texas 78712 (United States)

    1997-05-01

    The maximum normalized beta achieved in long-pulse tokamak discharges at low collisionality falls significantly below both that observed in short pulse discharges and that predicted by the ideal MHD theory. Recent long-pulse experiments, in particular those simulating the International Thermonuclear Experimental Reactor (ITER) [M. Rosenbluth {ital et al.}, {ital Plasma Physics and Controlled Nuclear Fusion} (International Atomic Energy Agency, Vienna, 1995), Vol. 2, p. 517] scenarios with low collisionality {nu}{sub e{sup {asterisk}}}, are often limited by low-m/n nonideal magnetohydrodynamic (MHD) modes. The effect of saturated MHD modes is a reduction of the confinement time by 10{percent}{endash}20{percent}, depending on the island size and location, and can lead to a disruption. Recent theories on neoclassical destabilization of tearing modes, including the effects of a perturbed helical bootstrap current, are successful in explaining the qualitative behavior of the resistive modes and recent results are consistent with the size of the saturated islands. Also, a strong correlation is observed between the onset of these low-m/n modes with sawteeth, edge localized modes (ELM), or fishbone events, consistent with the seed island required by the theory. We will focus on a quantitative comparison between both the conventional resistive and neoclassical theories, and the experimental results of several machines, which have all observed these low-m/n nonideal modes. This enables us to single out the key issues in projecting the long-pulse beta limits of ITER-size tokamaks and also to discuss possible plasma control methods that can increase the soft {beta} limit, decrease the seed perturbations, and/or diminish the effects on confinement. {copyright} {ital 1997 American Institute of Physics.}

  19. SST and ADITYA tokamak research in India

    International Nuclear Information System (INIS)

    Steady state operation of tokamaks plays an important role in high temperature magnetically confined plasma research. Steady state Superconducting Tokamak (SST) programme in India deals with the development of various technologies in this direction. SST-1 machine has been engineered and is being fabricated at the Institute for Plasma Research. The objectives of the machine are to study physics of plasma processes under steady state condition and develop the technologies related to steady state operation. Various sub-systems are being prototyped and developed. SST-1 is a large aspect ratio machine with a major radius of 1.1 m and a plasma minor radius of 0.2 m with elongation of 1.7 to 1.9 and triangularity of 0.5 to 0.7. It has been designed for 1000 sec operation at 3 T toroidal magnetic eld. Neutral beam Injection and Radio frequency heating systems are being developed to heat the plasma. Lower hybrid Current Drive system would sustain 200 kA of plasma current during 1000 sec operation. ADITYA tokamak has been upgraded with new diagnostics and RF heating systems. Thomson Scattering and ECE diagnostics have been operated. 200 kW Ion Cyclotron Resonance Heating (ICRH) and 200 kW Electron Cyclotron Resonance Heating (ECRH) systems have been successfully commissioned. RF assisted initial breakdown experiments have been initiated with these systems. (author)

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

    International Nuclear Information System (INIS)

    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

  1. Ferromagnetic and resistive wall effects on beta limit in a tokamak

    International Nuclear Information System (INIS)

    Ferromagnetic and resistive wall effect on beta limit in a tokamak is investigated. It is shown that the beta limit is reduced to 90% of that without ferromagnetic effect for high aspect ratio tokamak, if the ferromagnetic wall of relative permeability of 2 is used. The effect of toroidal plasma flow is also investigated, and the flow velocity of 0.03vta, vta is toroidal Alfven velocity, is sufficient for the resistive wall to have stability effect of ideal wall. Both the resistive wall and ideal kink modes are destabilized by the ferromagnetic wall effects. (author)

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

    International Nuclear Information System (INIS)

    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

  3. Production and maintenance of high poloidal beta tokamak plasmas by means of rf current drive

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-03-27

    It is shown that in tokamak plasmas sustained by rf current drive, the contribution of the suprathermal rf-driven electron population to the poloidal beta (..beta../sub p/) can be substantial if the total current is comparable to the Alfven critical current, I/sub A/ = (4..pi..mcv/..mu../sub 0/ec)..gamma... Equilibria with values of epsilon..beta../sub p/ up to approximately 1.3 were obtained, and no equilibrium or gross stability limits were observed.

  4. Magnetized plasma flow injection into tokamak and high-beta compact torus plasmas

    Science.gov (United States)

    Matsunaga, Hiroyuki; Komoriya, Yuuki; Tazawa, Hiroyasu; Asai, Tomohiko; Takahashi, Tsutomu; Steinhauer, Loren; Itagaki, Hirotomo; Onchi, Takumi; Hirose, Akira

    2010-11-01

    As an application of a magnetized coaxial plasma gun (MCPG), magnetic helicity injection via injection of a highly elongated compact torus (magnetized plasma flow: MPF) has been conducted on both tokamak and field-reversed configuration (FRC) plasmas. The injected plasmoid has significant amounts of helicity and particle contents and has been proposed as a fueling and a current drive method for various torus systems. In the FRC, MPF is expected to generate partially spherical tokamak like FRC equilibrium by injecting a significant amount of magnetic helicity. As a circumstantial evidence of the modified equilibrium, suppressed rotational instability with toroidal mode number n = 2. MPF injection experiments have also been applied to the STOR-M tokamak as a start-up and current drive method. Differences in the responses of targets especially relation with beta value and the self-organization feature will be studied.

  5. The Alfven wave spectrum of analytic high-beta tokamak equilibria

    International Nuclear Information System (INIS)

    This thesis addresses a number of problems regarding the equilibrium and stability of a tokamak plasma under fusion conditions. To get insight into the geometric effects on the behaviour of the most prominent global modes, a spectral study was carried out on a class of analytic, noncircular plasma equilibria. Parallel to this work, extending a previous high-beta energy principle a variational principle is developed for the numerical determination of the Alfven spectrum of a high-beta tokamak with arbitrary cross-section. Based on the Lagrangian formalism, representations were derived for the potential and kinetic energy in terms of arbitrary, non-orthogonal flux coordinates, which can be readily implemented in a numerical programme. The method is then tested by using the analytic equilibrium as input. 85 refs.; 32 figs.; 2 tabs

  6. Recent Activities on the Experimental Research Programme Using Small Tokamaks

    International Nuclear Information System (INIS)

    A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project (CRP) is discussed in this paper. Besides the presentation of the recent activities on the experimental research programme using small tokamaks and scientific results achieved at the participating laboratories, information is provided about the organisation of the co-ordinated research project. Future plans of the co-ordinated activities within the CRP are discussed

  7. Scaling of heat transport with beta in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Experiments in the DIII-D tokamak have measured the scaling of heat transport with beta (β) while all other dimensionless parameters are held constant for both H mode and L mode plasmas. Experimental results from the beta scaling of heat transport help to differentiate between various proposed mechanisms of turbulent transport. For L mode plasmas, the beta scaling of heat transport over the range 0.26≤βN≤0.49 is close to zero, with the thermal confinement time scaling as Bτth∝β-0.05+-0.10 and the effective (or one fluid) thermal diffusivity scaling as χeff∝χBβ0.11+-0.20. The beta scalings of the ion and electron thermal diffusivities are the same as the effective diffusivity to within the experimental errors. Higher values of beta are investigated in H mode plasmas, where a weak-to-moderate beta scaling of transport is observed over the range 0.8≤βN≤1.7, with the thermal confinement time scaling as Bτth∝β0.03+-0.11 and the effective thermal diffusivity scaling as χeff∝χBβ-0.54+-0.21. The ion channel is responsible for the favourable beta scaling of H mode plamas; the electron channel has no measurable beta dependence. These beta scalings determined by dimensionless parameter scans are much weaker than the predicted beta scalings from the L mode and H mode confinement scaling expressions that are currently being used to predict the performance of ITER. (author)

  8. Tokamak MHD Stability at High Beta and Low Plasma Rotation

    Science.gov (United States)

    Garofalo, A. M.; Reimerdes, H.; Lanctot, M. J.; Albrecht, J. T.; Okabayashi, M.; Solomon, W. M.; Jackson, G. L.; La Haye, R. J.; Strait, E. J.

    2006-10-01

    Recent high-beta DIII-D experiments with the new capability of balanced neutral beam injection show that the resistive wall mode (RWM) remains stable even with significant reductions in the neutral beam torque relative to pure co-injection. Previous DIII-D experiments showed a higher plasma rotation threshold (˜1-3%,A) for RWM stabilization when resonant magnetic braking was used to lower the plasma rotation. We speculate that the previously observed rotation threshold corresponds to the entrance into a forbidden band of rotation that results from torque balance including the resonant field amplification by the stable RWM. Previous and recent experimental data show a bifurcation taking place when the plasma rotation is reduced to half its unperturbed value, consistent with theory [1]. This hypothesis may have implications for both RWM stability and error field tolerances in ITER. 4pt[1] R. Fitzpatrick, Nucl. Fusion 33, 1049 (1993).

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

    International Nuclear Information System (INIS)

    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. Extremely fast vertical displacement event induced by a plasma {beta}{sub p} collapse in high {beta}{sub p} tokamak disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Yukiharu; Yoshino, Ryuji [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Pomphrey, N.; Jardin, S.C.

    1996-05-01

    In a vertically elongated ({kappa} {approx} 1.5), high {beta}{sub p}({beta}{sub p} {approx} 1.7) tokamak with a resistive shell, extremely fast vertical displacement events (VDE`s) induced by a model of strong {beta}{sub 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 {beta}{sub p} tokamak ({beta}{sub p} {approx} 0.2), was not observed during the VDE evolution, the observed growth rate of VDE`s was almost five times ({gamma} {approx} 655 sec{sup -1}) faster than the growth rate of the usual positional instability ({gamma} {approx} 149 sec{sup -1}). The essential mechanism of the {beta}{sub 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{sub 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 {beta}{sub p} collapse and the n-index of the equilibria just before the {beta}{sub p} collapse occurs. A new guide line for designing the fusion reactor is proposed with considering the impact of disruptions. (author).

  11. Research using small tokamaks. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The technical reports in these proceedings were presented at the IAEA Technical Committee Meeting on research Using Small Tokamaks, held in Ahmedabad, India, 6-7 December 1995. The purpose of this annual meeting is to provide a forum for the exchange of information on various small and medium sized plasma experiments, not only for tokamaks. The potential benefits of these research programmes are to: test theories, such as effects of the plasma rotation; check empirical scalings, such as density limits; develop fusion technology hardware; develop plasma diagnostics; such as tomography; and to train scientists, engineers, technicians, and students, particularly in developing IAEA Member States

  12. DIII-D Advanced Tokamak Research Overview

    International Nuclear Information System (INIS)

    This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously βNH of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues

  13. Spherical tokamak research for fusion reactor

    International Nuclear Information System (INIS)

    Between ITER and the commercial fusion reactor, there are many technological problems to be solved such as cost, neutron and steady-state operation. In the conceptual design of VECTOR and Slim CS reactors it was shown that the key is 'low aspect ratio'. The spherical tokamak (ST) has been expected as the base for fusion reactors. In US, ST is considered as a non-superconducting reactor for use in the neutron irradiation facility. Conceptual design of the superconducting ST reactor is conducted in Japan and Korea independently. In the present article, the prospect of the ST reactor design is discussed. (author)

  14. Measurement of high-beta tokamak pressure profiles with multipoint Thomson scattering

    International Nuclear Information System (INIS)

    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 1015 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 1014 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

  15. Role of explosive instabilities in high-$\\beta$ disruptions in tokamaks

    CERN Document Server

    Aydemir, A Y; Lee, S G; Seol, J; Park, B H; In, Y K

    2016-01-01

    Intrinsically explosive growth of a ballooning finger is demonstrated in nonlinear magnetohydrodynamic calculations of high-$\\beta$ disruptions in tokamaks. The explosive finger is formed by an ideally unstable n=1 mode, dominated by an m/n=2/1 component. The quadrupole geometry of the 2/1 perturbed pressure field provides a generic mechanism for the formation of the initial ballooning finger and its subsequent transition from exponential to explosive growth, without relying on secondary processes. The explosive ejection of the hot plasma from the core and stochastization of the magnetic field occur in Alfv\\'enic time scales, accounting for the extremely fast growth of the precursor oscillations and the rapidity of the thermal quench in some high-$\\beta$ disruptions.

  16. Finite beta effects on turbulent transport in tokamak plasmas

    International Nuclear Information System (INIS)

    The research on the transport properties of magnetically confined plasmas plays an essential role towards the achievement of practical nuclear fusion energy. An economically viable fusion reactor is expected to operate at high plasma pressure. This implies that the detailed study of the impact of electromagnetic effects, whose strength increases with increasing pressure, is of critical importance. In the present work, the electromagnetic effects on the particle, momentum and heat transport channels have been investigated, with both analytical and numerical calculations. Transport processes due to a finite plasma pressure have been identified, their physical mechanisms have been explained, and their contributions have been quantified, showing that they can be significant under experimentally relevant conditions.

  17. Abstracts of the International seminar 'Experimental possibilities of KTM tokamak and research programme'

    International Nuclear Information System (INIS)

    The International seminar 'Experimental possibilities of KTM tokamak and research programme' was held in 10-12 October 2005 in Astana city (Kazakhstan). The seminar was dedicated to problems of KTM tokamak commissioning. The Collection of abstracts comprises 45 papers

  18. A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

    OpenAIRE

    Mitsuru Kikuchi

    2010-01-01

    Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR) best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fund...

  19. Effect of the Shafranov shift and the gradient of $\\beta$ on intrinsic momentum transport in up-down asymmetric tokamaks

    CERN Document Server

    Ball, Justin; Lee, Jungpyo; Cerfon, Antoine J

    2016-01-01

    Tokamaks with up-down asymmetric poloidal cross-sections spontaneously rotate due to turbulent transport of momentum. In this work, we investigate the effect of the Shafranov shift on this intrinsic rotation, primarily by analyzing tokamaks with tilted elliptical flux surfaces. By expanding the Grad-Shafranov equation in the large aspect ratio limit we calculate the magnitude and direction of the Shafranov shift in tilted elliptical tokamaks. The results show that, while the Shafranov shift becomes up-down asymmetric and depends strongly on the tilt angle of the flux surfaces, it is insensitive to the shape of the current and pressure profiles (when specific experimental parameters are kept fixed). Next, local nonlinear gyrokinetic simulations of these MHD equilibria are performed with GS2, which reveal that the Shafranov shift can significantly enhance the momentum transport. However, to be consistent, the effect of $\\beta'$ (i.e. the radial gradient of $\\beta$) on the magnetic equilibrium was also included....

  20. An analytic determination of beta poloidal and internal inductance in an elongated tokamak from magnetic probe measurements

    International Nuclear Information System (INIS)

    Analytic calculations of the magnetic fields available to magnetic diagnostics are performed for tokamaks with circular and elliptical cross sections. The explicit dependence of the magnetic fields on the poloidal beta and internal inductances is sought. For tokamaks with circular cross sections, Shafranov's results are reproduced and extended. To first order in the inverse aspect ratio expansion of the magnetic fields, only a specific combination of beta poloidal and internal inductance is found to be measurable. To second order in the expansion, the measurements of beta poloidal and the internal inductance are demonstrated to be separable but excessively sensitive to experimental error. For tokamaks with elliptical cross sections, magnetic measurements are found to determine beta poloidal and the internal inductance separately. A second harmonic component of the zeroth order field in combination with the dc harmonic of the zeroth order field specifies the internal inductance. The internal inductance in hand, measurement of the first order, first harmonic component of the magnetic field then determined beta poloidal. The degeneracy implicit in Shafranov's result (i.e. that only a combination of beta poloidal and internal inductance is measurable for a circular plasma cross section) reasserts itself as the elliptic results are collapsed to their circular limits

  1. 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)

    Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (ne ∼ 1 - 5 x 1014 (cm-3), Te ∼ 4 - 10 (eV), Bt ∼ 0.2 - 0.4(T)). Carbon impurity light, mainly the strong lines of CII(4267A, emitted by the C+ ions) and CIII (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 Hα 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 τp is comparable with the plasma energy confinement time τE; 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

  2. Optimizing the beta limit in DIII-D advanced tokamak discharges

    International Nuclear Information System (INIS)

    Results are presented from comparisons of modeling and experiment in studies to assess the best choice of safety factor (q) profile, pressure profile and discharge shape for high beta, steady-state, noninductive advanced tokamak operation. This is motivated by the need for high qminβN to maximize the self-driven bootstrap current while maintaining high toroidal beta to increase fusion gain. Experiment and theory both show that increases in the achievable normalized beta (βN) can be obtained through broadening of the pressure profile and use of a symmetric double-null divertor shape. The general trend is for βN to decrease as the minimum q value (qmin) increases, but with a broadened pressure profile, βN = 4 is obtained with qmin ∼ 2 and qminβN increases with qmin. Modeling of equilibria with near 100% bootstrap current indicates that operation with βN ∼ 5 should be possible with a sufficiently broad pressure profile. (author)

  3. Passive stabilization of MHD instabilities at high {beta}n in the HBT-EP Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gates, D.A.

    1993-11-01

    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 {beta} with {beta} {proportional_to}

    /B{sup 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 {beta} is near the Troyon limit. The location of the wall required for significant stabilization is b/a {approximately}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.

  4. Recent advancement in research and planning toward high beta steady state operation in KSTAR

    International Nuclear Information System (INIS)

    The goal of Korean Superconducting Tokamak Advanced Research (KSTAR) research is to explore stable improved confinement regimes and technical challenge for superconducting tokamak operation and thus, to establish the basis for predictable high beta steady state tokamak plasma operation. To fulfil the goal, the current KSTAR research program is composed of three elements: 1) Exploration of anticipated engineering and technology for a stable long pulse operation of high beta plasmas including Edge Localized Mode (ELM) control with the low n (=1, 2) Resonant Magnetic Perturbation (RMP) using in-vessel control coils and innovative non-inductive current drives. The achieved long pulse operation up to ∼50s and fully non-inductive current drive will be combined in the future. Study of efficient heat exhaust will be combined with an innovative divertor design/operation. 2) Exploration of the operation boundary through establishment of true stability limits of the harmful MagnetoHydroDynamic (MHD) instabilities and confinement of the tokamak plasmas in KSTAR, making use of the lowest error field and magnetic ripple simultaneously achieved among all tokamaks ever built. The intrinsic machine error field has a long history of research as the source of MHD instabilities and magnetic ripple is known to be a cause of energy loss in the plasma. The achieved high beta discharges at βN ∼4 and stable discharges at q95 (∼2) will be further improved. 3) Validation of theoretical modeling of MHD instabilities and turbulence toward predictive capability of stable high beta plasmas. In support of these research goals, the state of the art diagnostic systems, such as Electron Cyclotron Emission Imaging (ECEI) system in addition to accurate profile diagnostics, are deployed not only to provide precise 2D/3D information of the MHD instabilities and turbulence but also to challenge unresolved physics problems such as the nature of ELMs, ELM-crash dynamics and the role of the core

  5. Alfvén Acoustic Channel for Ion Energy in High-Beta Tokamak Plasmas

    Science.gov (United States)

    Bierwage, Andreas; Aiba, Nobuyuki; Shinohara, Kouji

    2015-01-01

    When the plasma beta (ratio of thermal to magnetic pressure) in the core of a tokamak is raised to values of several percent, as required for a thermonuclear fusion reactor, continuous spectra of long-wavelength slow magnetosonic waves enter the frequency band occupied by continuous spectra of shear Alfvén waves. It is found that these two branches can couple strongly, so that Alfvén modes that are resonantly driven by suprathermal ions transfer some of their energy to sound waves. Since sound waves are heavily damped by thermal ion Landau resonances, these results reveal a new energy channel that contributes to the damping of Alfvénic instabilities and the noncollisional heating of bulk ions, with potentially important consequences for confinement and fusion performance.

  6. Alfvén acoustic channel for ion energy in high-beta tokamak plasmas.

    Science.gov (United States)

    Bierwage, Andreas; Aiba, Nobuyuki; Shinohara, Kouji

    2015-01-01

    When the plasma beta (ratio of thermal to magnetic pressure) in the core of a tokamak is raised to values of several percent, as required for a thermonuclear fusion reactor, continuous spectra of long-wavelength slow magnetosonic waves enter the frequency band occupied by continuous spectra of shear Alfvén waves. It is found that these two branches can couple strongly, so that Alfvén modes that are resonantly driven by suprathermal ions transfer some of their energy to sound waves. Since sound waves are heavily damped by thermal ion Landau resonances, these results reveal a new energy channel that contributes to the damping of Alfvénic instabilities and the noncollisional heating of bulk ions, with potentially important consequences for confinement and fusion performance. PMID:25615474

  7. Experimental exploration of profile control in the Princeton Beta Experiment-Modified (PBX-M) tokamak

    International Nuclear Information System (INIS)

    The experimental program of the Princeton Beta Experiment-Modified (PBX-M) tokamak [Phys. Fluids B2, 1271 (1990)] is directed toward tailoring plasma profiles to achieve greater stability and confinement and to gain access to the second stability region. Modification of the current density profile has been achieved with lower-hybrid current drive (LHCD), leading to a regime free of global magnetohydrodynamic modes, while raising the value of q(0) above unity. The diffusion of the fast electrons produced by LHCD has been examined using two-dimensional hard x-ray imaging. Ion Bernstein waves (IBW) have been used for ion heating: a preliminary analysis shows that ion heating was spatially localized and in agreement with theoretical calculations. Divertor biasing has modified the electric field inside the last closed surface, resulting in the formation of a transport barrier, which in turn has reduced the threshold power of neutral beam injection (NBI) for H-mode transition by 25%

  8. Future directions in fusion research: Super high field tokamaks

    International Nuclear Information System (INIS)

    Recent experimental results and advances in magnet engineering suggest that super high field, high aspect ratio tokamak devices could be a very efficient way to achieve burning plasma conditions and could open up a new area of research. Copper magnet devices with fields of 13 to 25 T at the plasma are considered. The super high field approach could also provide advantages for ETR and demonstration/commercial reactor concepts (magnetic fields at the plasma in the 8 to 13 T range)

  9. Overview of Physics Research on the TCV Tokamak

    Czech Academy of Sciences Publication Activity Database

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

    Geneve: International Atomic Energy Agency, 2008, 1-1-0/OV. [IAEA Fusion Energy Conference/22nd./. Geneva (CH), 13.10.2008-18.10.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * parallel flow Subject RIV: BL - Plasma and Gas Discharge Physics http://www-pub.iaea.org/MTCD/Meetings/fec2008pp.asphttp://www.fec2008.ch/preprints/ov_1-1.pdf

  10. Overview of physics research on the TCV tokamak

    Czech Academy of Sciences Publication Activity Database

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

    2009-01-01

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

  11. Romanian research in the field of Tokamak fusion reactors

    International Nuclear Information System (INIS)

    To re-create the conditions of the sun and stars for the production of fusion energy on earth, scientists most accomplish three major tasks. They have already passed the first task by achieving the necessary temperatures. In same cases, they have attained temperatures as high as 510 million degrees, 20 times more then the temperature at the center of the sun. Secondly, they need to demonstrate sustained reactions where substantial amounts of energy are produced. The third major milestone for fusion would be operation of a demonstration fusion power plant. Many different magnetic confined schemes have been studied. The one which is receiving the greatest attention in the international magnetic fusion energy programme is the tokamak concept, and represents actually the most advanced fusion devices. The advantage of fusion are: - abundant fuel supply; - no risk of a nuclear accident; - no air pollution; - no high-level nuclear waste; - no generation of weapons material. The present objectives and research priorities of the fusion community are: - continuation of ongoing research; - concept improvements; - long term technology. Our research programme in the field of tokamak fusion reactions is performed mainly in the frame of international cooperation with 'I.V. Kurchatov' Nuclear Fusion Institute from Moscow, Institute of Applied Mathematics from Grenoble, Research Center from Cadarache, 'Max-Planck' Institute for Plasma Physics from Garching at Munich and Columbia University from New York. The activities carried out under our programme are closely coordinated with those of the European Atomic Energy Community and are related to current problems concerning equilibrium, stability, transport and diagnostics of tokamak plasmas. Our results are mentioned in the International Atomic Energy Agency's World Survey of Activities in Controlled Fusion Research in 1997 and the European Community's Reports EUR FUR BRU from 1993 and 1996. (author)

  12. Analytic high-beta tokamak equilibria with poloidal-sonic flow and their stability

    International Nuclear Information System (INIS)

    Full text: In magnetically confined plasmas, equilibrium flows may suppress instability and turbulent transport to give rise to the formation of transport barriers or pedestals. In axisymmetric toroidal systems like tokamaks, both of the poloidal and toroidal components of flow are important for these phenomena. The ideal magnetohydrodynamic (MHD) equations for equilibria with flow reduce to the so-called generalized Grad-Shafranov (GS) equation and the Bernoulli law with five free functions of the magnetic flux in axisymmetric systems. These systems of a nonlinear partial differential equation (PDE) and a nonlinear algebraic equation can be solved numerically by iteration schemes when the PDE is elliptic in the whole region. Recently, a new reduced set of equilibrium equations has been derived for high-beta tokamaks with flow in the order of the poloidal-sound velocity. This model includes the effect of poloidal flow on both the magnetic flux and the pressure in higher-order terms of asymptotic expansions in terms of the inverse aspect ratio. We have found analytical solutions for this reduced set of equations for linear profiles of the lowest order free functions and perform qualitative studies of equilibria with sub- and super-poloidal-sonic flow. Analytical representations of equilibria with flow may be useful for benchmarks and initial guesses of iteration as well as for qualitative studies of equilibrium flows and stability analyses. In this presentation, we will also discuss the stability of these analytical equilibria. Equilibrium flow brings non-Hermitian properties to linear perturbations. In contrast to the static equilibria, the variational formulation does not give necessary and sufficient conditions, and the growth and damping can be non-exponential where mode analysis does not work. Due to these reasons, stability analysis of equilibria with flow must be done more carefully than that of static equilibria. Based on the formalism of the reduced MHD

  13. A brief overview of Tokamak fusion research

    International Nuclear Information System (INIS)

    Fusion, the nuclear engine that powers the sun and stars, has been pursued by scientists for decades as the ultimate source of energy. It promises an almost inexhaustible fuel supply with the oceans containing sufficient fusion fuel to outlast the expected life of the sun. Fusion is a process whose waste is inert and whose components know no geographical bounds. Scientists have pondered the laws governing the fusion process since the 1940's, and since the late 1950's laboratory devices have been constructed to test and further develop the theories. To achieve fusion, the joining of light atomic nuclei (as opposed to the splitting of heavy elements in the fission process), the natural tendency of the nuclei to repel each other due to their like electrical charges must be overcome. As the fusion takes place, some of the matter of the nuclei is converted to energy. In the stars fusion is accomplished largely by enormous gravitational forces. On earth the fusion fuel must be heated by other means to increase the energy of the particles to force them to fuse. Therein lies the challenge of fusion research - how to heat sufficient matter to hundreds of millions of degrees and contain it long enough for a controlled and sustained fusion reaction to take place. The method that presently shows the most promise is to contain a plasma (an ionized gas - the fourth state of matter) in a magnetic field while heating the plasma by means of high energy neutral particle beams or radio frequency waves

  14. A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

    Directory of Open Access Journals (Sweden)

    Mitsuru Kikuchi

    2010-11-01

    Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

  15. High-beta equilibria in tokamaks with pressure anisotropy and toroidal flow

    Science.gov (United States)

    Layden, B.; Hole, M. J.; Ridden-Harper, R.

    2015-12-01

    We extend previous analytical calculations of 2D high-β equilibria in order-unity aspect ratio tokamaks with toroidal flow to include pressure anisotropy, assuming guiding-center theory for a bi-Maxwellian plasma and the ideal MHD Ohm's law. Equilibrium solutions are obtained in the core region (which fills most of the plasma volume) and the boundary layer. We find that pressure anisotropy with p∥>p⊥ ( p∥Ωmin ) were previously found to suppress the field-free region (diamagnetic hole) that exists in static isotropic high-β equilibria. We find that all equilibrium solutions with pressure anisotropy suppress the diamagnetic hole. For the static case with a volume-averaged toroidal beta of 70%, plasmas with max (p∥/p⊥)>α1=1.07 have equilibrium solutions. We find that α1 decreases with increasing toroidal flow speed, and above the flow threshold Ωmin we find α1=1 , so that all p∥>p⊥ plasmas have equilibrium solutions. On the other hand, for p∥p⊥ , while the converse is true for p∥

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

    International Nuclear Information System (INIS)

    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 106 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 106 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 CO2 interferometry measurements

  17. Proposals for an influential role of small tokamaks in mainstream fusion physics and technology research

    International Nuclear Information System (INIS)

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confinement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies, which because of the risky nature cannot be done in large machines without preliminary studies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. All these tasks are well recognised and reflected in documents and understood by the large tokamak teams. Recent experimental results will be presented of contributions to mainstream fusion physics and technology research on small tokamaks involved in the IAEA Coordinated Research Project 'Joint Research using small tokamaks', started in 2004

  18. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    International Nuclear Information System (INIS)

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new

  19. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Lee, K. W.; Hwang, C. K.; Hong, B. G.; Hong, G. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-05-01

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new.

  20. Stabilization of ideal modes by resistive walls in tokamaks with plasma rotation and its effect on the beta limit

    International Nuclear Information System (INIS)

    It is shown that pressure-driven, ideal external modes in tokamaks can be fully stabilized by resistive walls when the plasma rotates at some fraction of the sound speed. For wall stabilized plasmas, there are two types of potentially unstable external modes: those which are nearly locked to the wall and those which rotate with the plasma. For the modes rotating with the plasma, the stabilizing effect of the wall increases when the wall is brought closer to the plasma, while, for the wall-locked modes, the stabilization improves with increasing wall distance. When the plasma rotates at some fraction of the sound speed, there is a window of stability to both the wall-locked and the rotating mode. This window closes when beta exceeds a new limit which can be significantly higher than the wall-at-infinity limit. The stabilization depends principally on the toroidal coupling to sound waves and is affected by ion Landau damping. Two dimensional stability calculations are presented to evaluate the gains in beta limit resulting from this wall stabilization for different equilibria and rotation speeds. In particular, results are shown for advanced tokamak configurations with bootstrap fractions of ≅ 100%. (author) 14 figs., 25 refs

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

    International Nuclear Information System (INIS)

    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

  2. Transport of energetic ions in MHD-active high-beta plasmas of spherical tokamaks

    International Nuclear Information System (INIS)

    It is shown that high β (β is the ratio of plasma pressure to the magnetic field pressure) may deteriorate the confinement of trapped energetic ions in spherical tokamaks (ST) during MHD events, such as sawtooth oscillations and internal reconnection events (IRE). This result indicates that moderate rather than very high β may be preferable in STs. (author)

  3. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    International Nuclear Information System (INIS)

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method

  4. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Hwan [Department of Nanoscale Semiconductor Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Hong, Suk-Ho [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); National Fusion Research Institute (NFRI), Daejeon 305-333 (Korea, Republic of); Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook, E-mail: joykang@hanyang.ac.kr [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of)

    2015-12-15

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.

  5. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method.

    Science.gov (United States)

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-12-01

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method. PMID:26724028

  6. PPPL tokamak program

    International Nuclear Information System (INIS)

    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. Overview of recent and current research on the TCV tokamak

    Science.gov (United States)

    S. Codathe TCV Team

    2013-10-01

    Through a diverse research programme, the Tokamak à Configuration Variable (TCV) addresses physics issues and develops tools for ITER and for the longer term goals of nuclear fusion, relying especially on its extreme plasma shaping and electron cyclotron resonance heating (ECRH) launching flexibility and preparing for an ECRH and NBI power upgrade. Localized edge heating was unexpectedly found to decrease the period and relative energy loss of edge localized modes (ELMs). Successful ELM pacing has been demonstrated by following individual ELM detection with an ECRH power cut before turning the power back up to trigger the next ELM, the duration of the cut determining the ELM period. Negative triangularity was also seen to reduce the ELM energy release. H-mode studies have focused on the L-H threshold dependence on the main ion species and on the divertor leg length. Both L- and H-modes have been explored in the snowflake configuration with emphasis on edge measurements, revealing that the heat flux to the strike points on the secondary separatrix increases as the X-points approach each other, well before they coalesce. In L-mode, a systematic scan of the auxiliary power deposition profile, with no effect on confinement, has ruled it out as the cause of confinement degradation. An ECRH power absorption observer based on transmitted stray radiation was validated for eventual polarization control. A new profile control methodology was introduced, relying on real-time modelling to supplement diagnostic information; the RAPTOR current transport code in particular has been employed for joint control of the internal inductance and central temperature. An internal inductance controller using the ohmic transformer has also been demonstrated. Fundamental investigations of neoclassical tearing mode (NTM) seed island formation by sawtooth crashes and of NTM destabilization in the absence of a sawtooth trigger were carried out. Both stabilizing and destabilizing agents

  8. Geodesic acoustic modes in a fluid model of tokamak plasma : the effects of finite beta and collisionality

    CERN Document Server

    Singh, Rameswar; Gurcan, Ozgur D; Hennequin, Pascale; Vermare, L; Morel, Pierre; Singh, Raghvendra

    2015-01-01

    Starting from the Braginskii equations, relevant for the tokamak edge region, a complete set of nonlinear equations for the geodesic acoustic modes (GAM) has been derived which includes collisionality, plasma beta and external sources of particle, momentum and heat. Local linear analysis shows that the GAM frequency increases with collisionality at low radial wave number $k_{r}$ and decreases at high $k_{r}$. GAM frequency also decreases with plasma beta. Radial profiles of GAM frequency for two Tore Supra shots, which were part of a collisionality scan, are compared with these calculations. Discrepency between experiment and theory is observed, which seems to be explained by a finite $k_{r}$ for the GAM when flux surface averaged density $\\langle n \\rangle$ and temperature $\\langle T \\rangle$ are assumed to vanish. It is shown that this agreement is incidental and self-consistent inclusion of $\\langle n \\rangle$ and $\\langle T \\rangle$ responses enhances the disagreement more with $k_r$ at high $k_{r}$ . So ...

  9. Preliminary Study of Ideal Operational MHD Beta Limit in HL-2A Tokamak Plasmas

    Science.gov (United States)

    Shen, Yong; Dong, Jiaqi; He, Hongda; D. Turnbull, A.

    2009-04-01

    Magnetohydrodynamic (MHD) n = 1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of βcN = 2.56 for a typical HL-2A discharge with a plasma current Ip = 0.245 MA, and the scaling of βcN ~constant is confirmed.

  10. Preliminary Study of Ideal Operational MHD Beta Limit in HL-2A Tokamak Plasmas

    Institute of Scientific and Technical Information of China (English)

    SHEN Yong; DONG Jiaqi; HE Hongda; A. D. TURNBULL

    2009-01-01

    Magnetohydrodynamic (MHD) n=1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of βcN=2.56 for a typical HL-2A discharge with a plasma current Ip=0.245 MA, and the scaling of βcN~constant is confirmed.

  11. 20 years of research on the Alcator C-Mod tokamak

    OpenAIRE

    Greenwald, Martin; Bader, A; Baek, S.; M. Bakhtiari; Barnard, H.; Beck, W.; Bergerson, W; Bespamyatnov, I; Bonoli, P.; Brower, D; Brunner, D.; Burke, W.; Candy, J.; Churchill, M; Cziegler, I.

    2014-01-01

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of crit...

  12. An experimental study of lower-hybrid wave driven high-poloidal beta plasmas in the Versator II tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Squire, J.P.

    1993-01-01

    High poloidal beta plasma equilibria have been produced by both toroidally asymmetric and symmetric lower-hybrid rf injection in the Versator II tokamak. (Here [beta][sub p] is the ratio of the plasma pressure to the poloidal magnetic field pressure, and [epsilon] = a/R[sub 0] is the inverse aspect ratio). In both cases the plasma current was fully sustained by the rf, with the loop voltage negligibly small. The rf-created high energy electron distribution function, which provides a large fraction of plasma current and pressure in these plasmas, is studied by means of X-ray spectroscopy of the electron-ion perpendicular to the toroidal magnetic field and emission measurements at a full range of angles to the toroidal magnetic field have been carried out. For low plasma current equilibria, [beta][sub p] is enhanced, and an outward shift in major radius of the X-ray emission profile peak was observed, corresponding to a Shafranov shift of the magnetic axis. The LHCD equilibrium current profiles was determined from the X-ray emission profile. The LHCD X-ray data indicate a highly anisotropic energetic electron distribution function with a density approximately 1% of the bulk electron density at the center. The stored energy of this distribution is much larger than the bulk energy and the global energy confinement time scales in agreement with the Kaye-Goldston L-mode scaling. At high values of [beta][sub p] [approximately] 3, LHH generates nearly the same plasma current as LHCD. The LHH equilibria has a reduction of up to a factor of four in the high energy X-ray emission flux, as compared to LHCD. Modeling of the electron distribution function in the LHH case indicates that only one third of the current is carried by the high energy electrons. At lower values of [beta][sub p] [approximately] 1.5, LHH requires the assistance of a small applied loop voltage to maintain the plasma current and the high energy electrons carry a majority of the plasma current.

  13. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D3He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  14. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  15. Electromagnetic stabilization of tokamak microturbulence in a high-$\\beta$ regime

    CERN Document Server

    Citrin, J; Goerler, T; Jenko, F; Mantica, P; Told, D; Bourdelle, C; Hatch, D R; Hogeweij, G M D; Johnson, T; Pueschel, M J; Schneider, M

    2014-01-01

    The impact of electromagnetic stabilization and flow shear stabilization on ITG turbulence is investigated. Analysis of a low-$\\beta$ JET L-mode discharge illustrates the relation between ITG stabilization, and proximity to the electromagnetic instability threshold. This threshold is reduced by suprathermal pressure gradients, highlighting the effectiveness of fast ions in ITG stabilization. Extensive linear and nonlinear gyrokinetic simulations are then carried out for the high-$\\beta$ JET hybrid discharge 75225, at two separate locations at inner and outer radii. It is found that at the inner radius, nonlinear electromagnetic stabilization is dominant, and is critical for achieving simulated heat fluxes in agreement with the experiment. The enhancement of this effect by suprathermal pressure also remains significant. It is also found that flow shear stabilization is not effective at the inner radii. However, at outer radii the situation is reversed. Electromagnetic stabilization is negligible while the flow...

  16. The Recent Research Work on the J-TEXT Tokamak

    International Nuclear Information System (INIS)

    Full text: The main results from the J-TEXT tokamak in the last two years, which emphasized the observation and analysis of MHD activity, are summarized and presented in this meeting. Static resonant magnetic perturbations generated by saddle coil currents are applied to J-TEXT Ohmic plasmas in order to study their influence on MHD instabilities. With sufficiently large RMPs, the m/n = 2/1 (m and n are the poloidal and toroidal mode numbers) mode locking is easily obtained. The analysis of the mode locking thresholds varied by scanning of the spatial phase of RMPs shows that the m/n = 2/1 component of intrinsic error field of the J-TEXT tokamak is about 0.4 Gs. In addition to normal mode locking events, the (partial) stabilization of the m/n = 2 / 1 tearing mode by moderate magnetic perturbation amplitude is observed experimentally. With experimental parameters as input, both the mode locking and mode stabilization by RMPs are also obtained from nonlinear numerical modeling based on reduced MHD equations. It is found that the suppression of the tearing mode by RMPs of moderate amplitude is possible for a sufficiently high plasma rotation frequency and low Alfven velocity. Gas puffing is also used to affect the MHD activity in J-TEXT. For example, neon gas injection can cause inverse sawtooth-like activity that spreads from the q = 1 surface to the axis; in particular, small amplitude m/n = 1 / 1 mode oscillations superimposed on the inverse sawtooth waveform around the q = 1 surface are observed after the impurity injection. Nevertheless, other impurities such as helium and argon impurities can't trigger such events. In addition, gas puffing can also be applied to mitigate disruptions, especially on the current quench phase. It is found that no runaway current generation occurs in intentionally provoked disruptions when the toroidal magnetic field is lower than 2.2 T. The runaway currents can be suppressed by the intensive gas puffing of H2 . To meet the

  17. Present status of fusion research: The next step tokamak (ITER) and the demonstration reaction (DEMO)

    International Nuclear Information System (INIS)

    In the search for new sources of energy, fusion offers great possibilities for the future with virtually inexhaustible reserves and a negligible basic fuel cost. On the other hand, the conditions for the thermonuclear reactions are difficult and complex to implement because of the temperature of approximately 100 million degrees necessary to initiate nuclear combustion. Research based on magnetic confinement where magnetic field are used to contain the electrically conducting plasma was initiated in the fifties. Among the wide variety of magnetic configurations studied to date, Tokamak-type devices have obtained the best experimental results and offer the best chances of obtaining a thermonuclear plasma within the next 15 years. The european strategy for achievement of the ultimate goal of construction of a prototype electricity generating reactor has been based on three major intermediate steps:(i) the present large Tokamak JET together with other specialized devices, to prove main aspects of scientific feasibility of fusion; (ii) the planned Next Step Tokamak, to complete demonstration of scientific feasibility and to establish a solid basis for the evaluation of the technological feasibility of fusion; (iii) a DEMO reactor to complete demonstration of technological feasibility and to establish a solid basis for the evaluation of the commercial feasibility of fusion. Plasma with thermonuclear parameters have been produced in several tokamaks, JET in particular. 4 figs., 1 tab., 9 refs. (author)

  18. Equilibrium and stability aspects of a screw-pinch based on the sharp-boundary model of a high-beta tokamak

    International Nuclear Information System (INIS)

    The sharp-boundary model of a high-beta tokamak surrounded by force-free currents (FFC) should yield a good description of the magnetohydroynamic stability of a screw-pinch: a tokamak with uniform q-profile. To arrive at the relationship between the equilibrium parameters giving rise to such a q-profile in the FFC region the poloidal field outside the plasma (with a prescribed cross-section) must be determined. An analytical solution has been derived for this field from which the desired relationship can be obtained by numerical means. From the results of a number of cross-sections it is evident that an approximation can be made that leads to this relationship more readily. Based on the latter the stability of a screw-pinch with a number of different cross-sections has been analyzed

  19. Results of Joint Experiments and other IAEA Activities on Research Using Small Tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Van Oost, G.; Peleman, P.; Brotánková, Jana; Dejarnac, Renaud; Dufková, Edita; Ďuran, Ivan; Hron, Martin; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zajac, Jaromír; Berni, L.A.; Del Bosco, E.; Ferreira, J.G.; Simões, J.R.; Berta, M.; Dunai, D.; Tál, B.; Zoletnik, S.; Malaquias, A.; Mank, G.; Figueiredo, H.; Kuznetsov, Y.; Ruchko, L.; Hegazy, H.; Ovsyannikov, A.; Sukhov, E.; Vorobjev, G.M.; Dreval, N.; Singh, A.; Budaev, V.; Kirnev, G.; Kirneva, N.; Kuteev, B.; Melnikov, A.; Nurov, D.; Sokolov, M.; Vershkov, V.; Khorshid, P.; Gonzales, R.; El Chama Neto, I.; Kraemer-Flecken, A.W.; Soldatov, V.; Brotas, B.; Carvalho, P. S.; Coelho, R.; Duarte, A.; Fernandes, H.; Figueiredo, J.; Fonseca, A.; Gomes, R.; Nedzelskiy, I.; Neto, A.; Ramos, G.; Santos, J.; Silva, C.; Valcárcel, D.; Gutierrez Tapia, C.R.; Krupnik, L.I.; Petrov, L.; Kolokoltsov, M.; Herrera, J.; Nieto-Perez, M.; Czarnecka, A.; Balan, P.; Sharnin, A.; Pavlov, V.

    Vienna : International Atomic Energy Agency, 2008, OV/P1-1-OV/P1-8. ISBN N. [IAEA Fusion Energy Conference/22nd./. Geneva (CH), 13.10.2008-18.10.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * joint experiment * turbulence * transport barrier * improvement confinement * electric field Subject RIV: BL - Plasma and Gas Discharge Physics http://www-pub.iaea.org/MTCD/Meetings/FEC2008/ov_p1-1.pdf

  20. Objectives and design of the JT-60 superconducting tokamak

    International Nuclear Information System (INIS)

    A fully superconducting tokamak named as JT-60SC is designed for the modification program of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-beta steady-state operation in the use of low radio-activation ferritic steel in low ν and ρ regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented. (author)

  1. RESEARCH OF BETA AS ADEQUATE RISK MEASURE-IS BETA STILL ALIVE?

    Directory of Open Access Journals (Sweden)

    Ante Perković

    2011-02-01

    Full Text Available The capital asset pricing model (CAPM is one of the most important models in financial economics and it has a long history of theoretical and empirical investigations. The main underlying concept of the CAPM model is that assets with a high risk (high beta should earn a higher return than assets with a low risk (low beta and vice versa. The implication which can be drawn out of this is that all assets with a beta above zero bear some risk and therefore their expected return is above the return of the risk-free rate. In this research observation on monthly stock prices on Croatian stock market from January 1st 2005 until December 31st 2009 is used to form our sample. CROBEX index is used as proxy of the market portfolio. The results demonstrate that beta can not be trusted in making investment decisions and rejects the validity of the whole CAPM model on Croatian stock market.

  2. Texas Experimental Tokamak: A plasma research facility. Technical progress report, November 1, 1993--October 31, 1994

    International Nuclear Information System (INIS)

    The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics in order to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks and in particular to understand the role of turbulence. So that they can continue to study the physics that is most relevant to the fusion program, TEXT completed a significant device upgrade this year. The new capabilities of the device and new and innovative diagnostics were exploited in all main program areas including: (1) configuration studies; (2) electron cyclotron heating physics; (3) improved confinement modes; (4) edge physics/impurity studies; (5) central turbulence and transport; and (6) transient transport. Details of the progress in each of the research areas are described

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

    International Nuclear Information System (INIS)

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

  4. Research tokamak system with multi-mode discharges using inverter power supply

    International Nuclear Information System (INIS)

    In Current Sustaining Tokamak in Nagoya university (CSTN)-IV research tokamak system using a compact 40kHz pulse width modulation (PWM) inverter power supply, which is controlled through LabVIEW program, we construct a new tokamak discharge system with multi-mode including a stable alternating current discharge and a high-repetition high-duty one. These discharge modes can be operated continuously for as long as 60sec. The continuous discharge with long duration is able to simulate the important physical and chemical processes of long time discharges in fusion devices, in which the heat load to the wall and the particle balance in the plasma-wall system are crucial topics in order to realize a long pulse fusion reactor, like ITER. Employing ergodic divertor (ED) is one of tools to control the particle balance and the heat load to the wall. In addition, we installed another inverter power supply to generate a rotating magnetic perturbation for dynamic ergodic divertor (DED) with the appropriate measurement system so that we may carry out experiments on heat and particle control with DED at long time operation. (author)

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

    International Nuclear Information System (INIS)

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

  6. Texas Experimental Tokamak, a plasma research facility: Technical progress report

    Energy Technology Data Exchange (ETDEWEB)

    Wootton, A.J.

    1995-08-01

    In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively.

  7. Texas Experimental Tokamak, a plasma research facility: Technical progress report

    International Nuclear Information System (INIS)

    In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively

  8. Results of Joint Experiments and other IAEA activities on research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Brotánková, Jana; Dejarnac, Renaud; Dufková, Edita; Ďuran, Ivan; Hron, Martin; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zajac, Jaromír

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104026-104026. ISSN 0029-5515. [IAEA Fusion Energy Conference/22nd./. Geneva, 13.10.2008-18.10.2008] R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * probe diagnostics * sheared flows * edge plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://iopscience.iop.org/0029-5515/49/10/104026

  9. Korea Superconducting tokamak advanced research project - Development of heating system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byung Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The heating and current drive systems for KSTAR based on multiple technologies (neutral beam, ion cyclotron, lower hybrid and electron cyclotron) have been designed to provide heating and current drive capabilities as well as flexibility in the control of current density and pressure profiles needed to meet the mission and research objectives of the machine. They are designed to operate for long-pulse lengths of up to 300 s. The NBI system initially delivers 8 MW of neutral beam power to the plasma from one co-directed beam line and shall be upgraded to provide 20 MW of neutral beam power with two co-directed beam lines plus one counter-directed beam line. It will be capable of being reconfigured such that the source arrangement is changed from horizontal to vertical stacking, with 6 MW beam power to the plasmas per beam line, in order to facilitate profile control. The RF system initially delivers 6 MW of rf power to the plasma, using a single four-strap antenna mounted in a midplane port. The system will be upgraded to proved 12 MW of rf power through 2 adjacent ports. In the first phase, we completed the basic design of RF system and the system have the capabilities to be operationable for pulse length up to 300 sec and in the 25-60 MHz frequency range. Lower hybrid system initially provides 1.5 MW LH rf power to the plasma at 3.7 GHz through a horizontal port, which has a capability to be operated for pulse length up to 300 sec, and shall be upgraded to provide 4.5 MW of LH rf power to the plasma. In the first phase, we completed the basic design of LHCD system which incorporate the TPX-type launcher and independently phase-changeable transmission system for the fully phased coupler. The ECH system will deliver up to 0.5 MW of power to the plasma for up to 0.5 sec. In the first phase, we completed the basic design of ECH system which includes an 84 GHz gyrotron system, a transmission system, and a launcher. The basic design of the low loss transmission system

  10. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research.

    Science.gov (United States)

    Lampert, M; Anda, G; Czopf, A; Erdei, G; Guszejnov, D; Kovácsik, Á; Pokol, G I; Réfy, D; Nam, Y U; Zoletnik, S

    2015-07-01

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera's measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties. PMID:26233377

  11. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    International Nuclear Information System (INIS)

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties

  12. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    Energy Technology Data Exchange (ETDEWEB)

    Lampert, M. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); BME NTI, Budapest (Hungary); Anda, G.; Réfy, D.; Zoletnik, S. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); Czopf, A.; Erdei, G. [Department of Atomic Physics, BME IOP, Budapest (Hungary); Guszejnov, D.; Kovácsik, Á.; Pokol, G. I. [BME NTI, Budapest (Hungary); Nam, Y. U. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-07-15

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

  13. Evidence for the Importance of Trapped Particle Resonances for Resistive Wall Mode Stability in High Beta Tokamak Plasmas

    International Nuclear Information System (INIS)

    Active measurements of the plasma stability in tokamak plasmas reveal the importance of kinetic resonances for resistive wall mode stability. The rotation dependence of the magnetic plasma response to externally applied quasistatic n=1 magnetic fields clearly shows the signatures of an interaction between the resistive wall mode and the precession and bounce motions of trapped thermal ions, as predicted by a perturbative model of plasma stability including kinetic effects. The identification of the stabilization mechanism is an essential step towards quantitative predictions for the prospects of ''passive'' resistive wall mode stabilization, i.e., without the use of an ''active'' feedback system, in fusion-alpha heated plasmas.

  14. The CIT [compact ignition tokamak] pellet injection system: Description and supporting research and development

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) will use an advance, high-velocity pellet injection system to achieve and maintain ignited plasmas. Two pellet injectors are provided: a moderate-velocity (1-to 1.5-km/s), single-stage pneumatic injector with high reliability and a high-velocity (4- to 5-km/s), two-stage pellet injector that uses frozen hydrogenic pellets encased in sabots. Both pellet injectors are qualified for operation with tritium feed gas. Issues such as performance, neutron activation of injector components, maintenance, design of the pellet injection vacuum line, gas loads to the reprocessing system, and equipment layout are discussed. Results and plans for supporting research and development (R and D) in the areas of tritium pellet fabrication and high-velocity, repetitive two-stage pneumatic injectors are presented. 7 refs., 4 figs., 2 tabs

  15. Remote maintenance design activities and research and development accomplishments for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    The use of deuterium-tritium (D-T) fuel for the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations. The remote operations consist of removing and replacing such components as first wall armor protection tiles, radio-frequency (rf) heating modules, and diagnostic modules. The major pieces of equipment being developed for maintenance activities internal to the vacuum vessel include an articulated boom manipulator (ABM), an inspection manipulator, and special tooling. For activities external to the vessel, the equipment includes a bridge-mounted manipulator system, decontamination equipment, hot cell equipment, and solid radiation-waste (rad-waste) handling and packaging equipment. The CIT Project is completing the conceptual design phase; research and development (R and D) activities, which include demonstrations of remote maintenance operations on full-size partial mock-ups are under way. 5 figs

  16. Researches on the Neutral Gas Pressure in the Divertor Chamber of the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    WANGMingxu; LIBo; YANGZhigang; YANLongwen; HONGWenyu; YUANBaoshan; LIULi; CAOZeng; CUIChenghe; LIUYong; WANGEnyao; ZHANGNianman

    2003-01-01

    The neutral gas pressure in divertor chamber is a very basic and important physics parameter because it determines the temperature of charged particles, the thermal flux density onto divertor plates, the erosion of divertor plates, impurity retaining and exhausting, particle transportation and confinement performance of plasma in tokamaks. Therefore, the pressure measurement in divertor chamber is taken into account in many large tokamaks.

  17. The response of short-scale density fluctuations to the activity of beta-induced Alfvén eigenmodes during strong tearing modes on EAST tokamak

    Science.gov (United States)

    Cao, G. M.; Li, Y. D.; Li, Q.; Sun, P. J.; Wu, G. J.; Hu, L. Q.; the EAST Team

    2015-08-01

    Beta-induced Alfvén eigenmodes (BAEs) during strong tearing modes (TMs) have been frequently observed in fast-electron plasmas of EAST tokamak. The dynamics of the short-scale ({k}\\perp {ρ }s~{1.5-4.3}) density fluctuations during the activity of BAEs with strong TMs has been preliminarily investigated by a tangential CO2 laser collective scattering system. The results suggest the active, but different, response of short-scale density fluctuations to the TMs and BAEs. In the low-frequency (0-10 kHz) part of density fluctuations, there are harmonic oscillations totally corresponding to those of TMs. In the medium-high frequency (10-250 kHz) part of density fluctuations, with the appearance of the BAEs, the medium-high frequency density fluctuations begin to be dominated by several quasi-coherent (QC) modes, and the frequencies of the QC modes seem to be related to the changes of both TMs and BAEs. These results would shed some light on the understanding of the multi-scale interaction physics.

  18. First results on disruption mitigation by massive gas injection in Korea Superconducting Tokamak Advanced Research

    International Nuclear Information System (INIS)

    Massive gas injection (MGI) system was developed on Korea Superconducting Tokamak Advanced Research (KSTAR) in 2011 campaign for disruption studies. The MGI valve has a volume of 80 ml and maximum injection pressure of 50 bar, the diameter of valve orifice to vacuum vessel is 18.4 mm, the distance between MGI valve and plasma edge is ∼3.4 m. The MGI power supply employs a large capacitor of 1 mF with the maximum voltage of 3 kV, the valve can be opened in less than 0.1 ms, and the amount of MGI can be controlled by the imposed voltage. During KSTAR 2011 campaign, MGI disruptions are carried out by triggering MGI during the flat top of circular and limiter discharges with plasma current 400 kA and magnetic field 2–3.5 T, deuterium injection pressure 39.7 bar, and imposed voltage 1.1–1.4 kV. The results show that MGI could mitigate the heat load and prevent runaway electrons with proper MGI amount, and MGI penetration is deeper under higher amount of MGI or lower magnetic field. However, plasma start-up is difficult after some of D2 MGI disruptions due to the high deuterium retention and consequently strong outgassing of deuterium in next shot, special effort should be made to get successful plasma start-up after deuterium MGI under the graphite first wall.

  19. The Fusion Science Research Plan for the Major U.S. Tokamaks. Advisory report

    International Nuclear Information System (INIS)

    In summary, the community has developed a research plan for the major tokamak facilities that will produce impressive scientific benefits over the next two years. The plan is well aligned with the new mission and goals of the restructured fusion energy sciences program recommended by FEAC. Budget increases for all three facilities will allow their programs to move forward in FY 1997, increasing their rate of scientific progress. With a shutdown deadline now established, the TFTR will forego all but a few critical upgrades and maximize operation to achieve a set of high-priority scientific objectives with deuterium-tritium plasmas. The DIII-D and Alcator C-Mod facilities will still fall well short of full utilization. Increasing the run time in vii DIII-D is recommended to increase the scientific output using its existing capabilities, even if scheduled upgrades must be further delayed. An increase in the Alcator C-Mod budget is recommended, at the expense of equal and modest reductions (~1%) in the other two facilities if necessary, to develop its capabilities for the long-term and increase its near-term scientific output.

  20. The power-supply control system for the toroidal magnets of the JFT-2M nuclear fusion research tokamak

    International Nuclear Information System (INIS)

    Mitsubishi Electric has completed the control system for the flywheel-equipped DC motor-generator that powers the toroidal field coils of the JFT-2M nuclear fusion research Tokamak at the Japan Atomic Energy Research Institute. The motor-generator, which has a maximum pulse output of 2.7 kV and 19 kA, was introduced in the May '96 issue of Giho. This article reports the key features of the coil current control. Emphasis is placed on the motor-generator field control provided by the thyristor converters, which opens the contacts of the main circuit switches in addition to controlling coil current. (author)

  1. Varennes Tokamak

    International Nuclear Information System (INIS)

    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

  2. Bootstrap current in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kessel, C.E.

    1994-03-01

    The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and {beta}{sub p} must be kept below a critical value.

  3. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

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

  4. 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.

  5. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  6. High Beta Steady State Research and Future Directions on JT-60U and JFT-2M

    Science.gov (United States)

    Ishida, Shinichi

    2003-10-01

    JT-60U and JFT-2M research is focused on high beta steady state operation towards economically and environmentally attractive reactors. In JT-60U, a high-βp H-mode plasma was sustained with βN 2.7 for 7.4 s in which neoclassical tearing modes (NTMs) limited the attainable β_N. Real-time tracking NTM stabilization system using ECCD demonstrated complete suppression of NTM leading to recovery of βN before onset of NTM. Performance in a fully non-inductive H-mode plasma was improved up to n_i(0) τE T_i(0) = 3.1 x 10^20 keV s m-3 using N-NBCD with βN 2.4, HH_y,2=1.2 and bootstrap fraction f_BS 0.5. ECH experiments extended the confinement enhancement for dominantly electron heated reversed shear plasmas up to HH_y,2 2 at T_e/Ti 1.25. A world record ECCD efficiency, 4.2 x 10^18 A/W/m^2, was achieved at Te 23 keV with a highly localized central current density. Innovative initiation and current build-up without center solenoid currents were established by LHCD/ECH and bootstrap current up to f_BS 0.9. In JFT-2M, the inside of the vacuum vessel wall was fully covered with low-activation ferritic steel plates to investigate their use in plasmas near fusion conditions. High βN plasmas were produced up to βN = 3.3 with an internal transport barrier (ITB) and a steady H-mode edge. A new H-mode regime with steady high recycling (HRS) and an ITB was exploited leading to βN H_89P 6.2 at n_e/nG 0.7. In 2003, JT-60U will be able to operate for the duration up to 65 s at 1 MA/2.7 T and the heating/current-drive duration up to 30 s at 17 MW to prolong high-βN and/or high-f_BS discharges with feedback controls. JFT-2M is planning to implement wall stabilization experiments in 2004 to pursue plasmas above the ideal no-wall limit using a ferritic wall. The modification of JT-60 to a fully superconducting tokamak is under discussion to explore high-β steady state operation in collision-less plasmas well above no-wall limit with ferritic wall in a steady state.

  7. Planning for US ion cyclotron heating research relevant to the Compact Ignition Tokamak and Alcator C-Mod

    International Nuclear Information System (INIS)

    Ion cyclotron heating (ICH) has been chosen as the primary method for providing auxiliary heating power to the plasma in the Compact Ignition Tokamak (CIT). Sustained progress in ion cyclotron range of frequencies (ICRF) heating experiments, together with supporting technology development, continues to justify selection of this technique as the preferred one for heating CIT to ignition. However, the CIT requirements are sufficiently different from existing achievements that continued experimentation and development are needed to meet the goals of the CIT experiment with a high degree of reliability. The purpose of this report is fourfold: (1) to review briefly the physics and technology research and development (R and D) needs for ICH on CIT, (2) to review the status of and planned programs for ICH on US and international machines, (3) to propose a unified ''mainline'' R and D program specifically geared to testing components for CIT, and (4) to assess the needs for experiments including C-Mod, the Tokamak Fusion Test Reactor (TFTR), and DIII-D to provide earlier information and improved probability of success for CIT ICH. 4 refs., 4 figs., 5 tabs

  8. Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

    International Nuclear Information System (INIS)

    It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the Dα amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase

  9. Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research.

    Science.gov (United States)

    Lee, W; Park, H K; Lee, D J; Nam, Y U; Leem, J; Kim, T K

    2016-04-01

    The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm(-1). The upper limit corresponds to the normalized wavenumber kθρe of ∼0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed. PMID:27131668

  10. Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research

    Science.gov (United States)

    Lee, W.; Park, H. K.; Lee, D. J.; Nam, Y. U.; Leem, J.; Kim, T. K.

    2016-04-01

    The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm-1. The upper limit corresponds to the normalized wavenumber kθρe of ˜0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed.

  11. IPP Prague contributions to the IAEA technical committee meeting on research using small tokamaks

    International Nuclear Information System (INIS)

    The report contains four papers dealing with the results of the CASTOR tokamak experiments achieved during the 1992 campaign. In the first paper the results of correlation analysis of plasma density fluctuations are reported and the role of edge plasma fluctuations in the global plasma confinement is discussed. The subject of the next paper is the improved particle confinement observed in the lower hybrid current drive and in the limiter biasing experiments. In the third paper, the close connection between the magnetic fluctuations level and the intensity of hard X radiation produced by runaway electrons is pointed out. In the last paper the atomic beam source ARALLIS used for CASTOR plasma diagnostics is described and the results of its stand testing are presented. (J.U.)

  12. Solenoid-free plasma start-up in spherical tokamaks

    Science.gov (United States)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

  13. Solenoid-free plasma start-up in spherical tokamaks

    International Nuclear Information System (INIS)

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid. (topical review)

  14. DIII-D research operations

    International Nuclear Information System (INIS)

    The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R ampersand D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma

  15. Annual report of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development for the period of April 1, 1977 to March 31, 1978

    International Nuclear Information System (INIS)

    Research and development works in fiscal year 1977 of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development are described. 1) Theoretical studies on tokamak confinement have continued with more emphasis on computations. A task was started of developing a computer code system for mhd behavior of tokamak plasmas. 2) Experimental studies of lower hybrid heating up to 140 kW were made in JFT-2. The ion temperature was increased by 50% -- 60% near the plasma center. Plasma-wall interactions (particle and thermal fluxes to the wall, and titanium gettering) were studied. In JFT-2a (DIVA) ion sputtering, arcing and evaporation were identified, and the impurity ion sputtering was found to be a dominant origin of metal impurities in the present tokamaks. High temperature and high-density plasma divertor actions were demonstrated; i.e. the divertor decreases the radiation power loss by a factor of 3 and increases the energy confinement time by a factor of 2.5. Various diagnostic instruments operated sufficiently to provide useful information for the research with JFT-2 and JFT-2a(DIVA). 3) JFT-2 and JFT-2a(DIVA) operated as scheduled. Technological improvements were made such as titanium coating of the chamber wall, discharge cleaning and pre-ionization. 4) Detailed design of the prototype JT-60 neutral beam injector was made. A 200 kW, 650 MHz radiofrequency heating system for JFT-2 was completed; a lower hybrid heating experiment in JFT-2 was successful 5) In particle-surface interactions, the sputtering and surface erosion were studied. 6) Improvement designs of a superconducting cluster test facility and a test module coil were made in the toroidal coil development. 7) Second preliminary design of the tokamak experimental fusion reactor JXFR started in April 1977. Safety analyses were made of the main components and system of JXFR on the basis of the first preliminary design. (J.P.N.)

  16. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  17. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    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

  18. Experimental and modeling researches of dust particles in the HL-2A tokamak

    International Nuclear Information System (INIS)

    The investigation of dust particle characteristics in fusion devices has become more and more imperative. In the HL-2A tokamak, the morphologies and compositions of dust particles are analyzed by using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDX) with mapping. The results indicate that the sizes of dust particles are in a range from 1 μm to 1 mm. Surprisingly, stainless steel spheres with a diameter of 2.5 μm–30 μm are obtained. The production mechanisms of dust particles include flaking, disintegration, agglomeration, and arcing. In addition, dynamic characteristics of the flaking dust particles are observed by a CMOS fast framing camera and simulated by a computer program. Both of the results display that the ion friction force is dominant in the toroidal direction, while the centrifugal force is crucial in the radial direction. Therefore, the visible dust particles are accelerated toriodally by the ion friction force and migrated radially by the centrifugal force. The averaged velocity of the grain is on the order of ∼ 100 m/s. These results provide an additional supplement for one of critical plasma-wall interaction (PWI) issues in the framework of the International Thermonuclear Experimental Reactor (ITER) programme. (paper)

  19. Experimental and modeling researches of dust particles in the HL-2A tokamak

    Institute of Scientific and Technical Information of China (English)

    黄治辉; 严龙文; 冨田幸博; 冯震; 程钧; 洪文玉; 潘宇东; 杨青巍; 段旭如

    2015-01-01

    The investigation of dust particle characteristics in fusion devices has become more and more imperative. In the HL-2A tokamak, the morphologies and compositions of dust particles are analyzed by using a scanning electron microscopy (SEM) and an energy dispersive x-ray spectroscopy (EDX) with mapping. The results indicate that the sizes of dust particles are in a range from 1 µm to 1 mm. Surprisingly, the stainless steel spheres with a diameter of 2.5 µm–30 µm are obtained. Production mechanism of the dust particles includes flaking, disintegration, agglomeration, and arcing. In addition, dynamic characteristics of the flaking dust particles are observed by a CMOS fast framing camera and simulated by a computer program. Both of the results display that the ion friction force is dominant in the toroidal direction, while the centrifugal force is crucial in the radial direction. Therefore, the visible dust particles are accelerated toriodally by the ion friction force and migrated radially by the centrifugal force. The averaged velocity of the grain is on the order of∼100 m/s. These results provide an additional supplement for one of critical plasma-wall interaction (PWI) issues in the framework of International Thermonuclear Experimental Reactor (ITER) programme.

  20. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  1. The ETE spherical Tokamak project

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  2. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  3. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

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

  4. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    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)

  5. The life test of a DC circuit breaker of tokamak device JT-60 for a nuclear fusion research

    International Nuclear Information System (INIS)

    In the Tokamak devices for nuclear fusion research, the construction of the current transformer circuits having plasma as the secondary circuit and the change of the primary circuit current are necessary for generating current in the plasma. This is considered to be fairly difficult in practice if conventional methods using capacitor discharge and iron core coils are employed. Considering such circumstances, it was decided for JT-60 to use an air-core current transformer coil and to employ the method of storing energy in the form of current in the coil inductance instead of a capacitor. For this reason, a DC circuit breaker is required to interrupt coil current. The authors improved an AV vacuum breaker, which had been developed as the vacuum breaker of longitudinal magnetic field type applying a magnetic field in parallel with an arc, to get the one for DC circuit for the purpose of applying it to JT-60. In this paper, the operational characteristic of the DC breaker is described, the construction and function of the life test circuit is explained, and the test results are reported. Finally, interruptions of 10,000 times at 20 kA were carried out. It is successful that the restrike of arc occurring during tens of milli-seconds after interruptions was improved to 0.05% or less for 10,000 times operations. Further, it was found that the generation of arc restrike can be reduced practically to zero with two breakers in series. (Wakatsuki, Y.)

  6. Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. Y.; Hong, J. H.; Jang, J. H.; Park, J. S.; Choe, Wonho, E-mail: wchoe@kaist.ac.kr [Department of Physics, Korea Advanced Institute of Science and Technology (KAIST), 34141 Daejeon (Korea, Republic of); Impurity and Edge Plasma Research Center, KAIST, 34141 Daejeon (Korea, Republic of); Hahn, S. H.; Bak, J. G.; Lee, J. H.; Ko, W. H.; Lee, K. D.; Lee, S. H.; Lee, H. H.; Juhn, J.-W.; Kim, H. S.; Yoon, S. W.; Han, H. [National Fusion Research Institute, 34133 Daejeon (Korea, Republic of); Ghim, Y.-C. [Deparment of Nuclear and Quantum Engineering, KAIST, 34141 Daejeon (Korea, Republic of)

    2015-12-15

    It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the D{sub α} amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase.

  7. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma

  8. Tokamak burn control

    International Nuclear Information System (INIS)

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  9. Experimental investigation of m/n = 1/1 and high-order harmonic modes during the sawtooth oscillation in a low beta tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Xu Xiaoyuan; Wang Jun; Wen Yizhi; Yu Yi; Liu Adi; Lan Tao; Yu Changxuan [CAS Key Laboratory of Plasma Physics, Department of Modern Physics, University of Science and Technology of China, Hefei 230026 (China); Wan Baonian; Gao, Xiang; Sun Youwen [Institute of Plasma Physics, The Chinese Academy of Sciences, Hefei 230031 (China); Luhmann, N C Jr; Domier, C W; Xia, Z G; Shen Zuowei, E-mail: xyxu3@mail.ustc.edu.c [Department of Applied Science, University of California, Davis, California 95616 (United States)

    2010-01-15

    Sawtooth oscillations were investigated using an electron cyclotron emission imaging diagnostic technique on the HT-7 tokamak. High-order harmonic modes are observed in sawtooth precursors. They cause sharp pressure points, leading to the occurrence of reconnection events at more than one place, and which are not preferential on the low field side of the q{approx}1 radius at low density. The reconnection proceeds in two stages. In the first stage, a weak reconnection happens which is characterized by a slow emergence of a small amount of heat and particles through the finite openings. Subsequently, in the second stage, most of the remaining heat and particles in the core escape outward. The observations at low density indicate that high-order harmonic modes are responsible for the secondary reconnection stage, which has seldom been mentioned in reconnection model. The characteristics of the modes are then described by investigating a large number of sawteeth oscillations with different densities and q{sub a} values. It is found that when the density increases above a threshold at q{sub a} = 3.3, the m/n = 1/1 mode amplitude significantly exceeds its harmonics. The threshold decreases as q{sub a} increases.

  10. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-βN tokamak reactors. (author)

  11. High performance operational limits of tokamak and helical systems

    International Nuclear Information System (INIS)

    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)

  12. Beta decay measurements with ultracold neutrons: a review of recent measurements and the research program at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    We present a review of the motivation and results of recent experiments which utilize ultracold neutrons for measurements of neutron beta decay. Because these experiments hinge critically on the available ultracold neutron source technology, we also review the status of ultracold neutron source development, emphasizing the Los Alamos ultracold neutron facility and the ongoing beta decay research program sited there. (paper)

  13. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  14. Power and particle exhaust in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER`s nominal design positions; important directions for further research are identified.

  15. Research on High Pressure Gas Injection As a Method of Fueling, Disruption Mitigation and Plasma Termination for Future Tokamak Reactors

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    High-pressure gas injection has proved to be an effective disruption mitigation technique in DⅢ-D tokamak experiments. If the method can be applied in future tokamak reactors not only for disruption mitigation but also for plasma termination and fueling, it will have an attractive advantage over the pellet and liquid injection from the viewpoint of economy and engineering design. In order to investigate the feasibility of this option, a study has been carried out with relevant parameters for conveying tubes of different geometrical sizes and for different gases.These parameters include pressure drop, lagger time after the valve's opening, gas diffusion in an ultra-high vacuum condition, and particle number contour.

  16. OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    BURRELL,KH

    2002-11-01

    OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, the authors have made significant progress in developing the building blocks needed for AT operation: (1) the authors have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {le} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. They have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiation power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet

  17. The upgradation of Aditya Tokamak

    International Nuclear Information System (INIS)

    Aditya Tokamak is the first Indian tokamak, indigenously built and commissioned at the Institute for Plasma Research, Gandhinagar, Gujarat, India, in September, 1989. Aditya Tokamak has been in operation since more than 25 years. More than 30,000 discharges are taken and a large number of experiments are carried out, with plasma current ranging from 50 KA to 150 KA, lasting for 100 to 250 milliseconds. Various types of wall conditioning techniques and different hot plasma diagnostics are tested and operated on Aditya Tokamak. The experiments for turbulent particle transport and turbulence in the edge plasma, gas puffing, lithium coating, mitigation, plasma disruption, limiter and electron biasing, runaway discharges etc. led to many interesting results contributing immensely to the world of thermonuclear fusion. Experiments on Pre-ionization and Plasma heating by ICRH and ECRH are also worked out. The scientific objectives of Aditya tokamak Upgrade include Low loop voltage plasma start-up with strong pre-ionization having a good plasma control system. The upgrade is designed keeping in mind the experiments, disruption mitigation studies relevant to future fusion devices, runway mitigation studies, demonstration of Radio-frequency heating and current drive etc. This upgraded Aditya tokamak will be used for basic studies on plasma confinement and scaling to larger devices, development and testing of new diagnostics etc. This machine will be easily accessible compared to SST-1 and will be very useful for generation of technical and scientific expertise for future fusion devices. In this paper, especial features of the upgrade including various aspects of designing of new components for Aditya Upgrade tokamak is presented

  18. JT-60SA project for JA-EU broader approach satellite tokamak and national centralized tokamak

    International Nuclear Information System (INIS)

    JT-60 Super Advanced (JT-60SA) project is the joint project of ITER satellite tokamak by Japan and EU with Japanese Tokamak. The background, objects, device design, management of JT-60SA is stated. It consists of six chapters: the first chapter describes introduction, the second chapter states the objects of tokamak device complementing ITER, the third chapter contains research subjects and device performance such as plasma performance and demand for devices, operation scenario, control of MHD instability, and control of heat and particles, the forth chapter design of devices, the fifth chapter management and the sixth conclusion. In order to realize prototype reactor, improvement research of tokamak, development of reactor engineering technology, fusion reactor researches, tokamak theory and simulation, and social and environment safety research has to be advanced. (S.Y.)

  19. Tore Supra tokamak

    International Nuclear Information System (INIS)

    This part of the electricity uses chapter of the Engineers Techniques collection is entirely devoted to the technical description of Tore Supra tokamak. A thermonuclear fusion device with magnetic confinement control such as Tore Supra concentrates a huge amount of high power electro-technical and electronic equipments. These power systems play a major role and are sometimes boosted to their extreme limits. From these equipments we can find: big superconducting magnets, big cooled copper magnets, high-voltage power supplies with thyristors (320 MVA installed), several MW hyper-frequency sources, several MW accelerated atom injectors, cryogenic, heat extraction, high-vacuum pumping systems, etc.. The components developed for these applications are numerous and frequently original: superconductor for variable magnetic field, DC static circuit breaker with high switch-off capability (0.7 GVA), 2 MW tetrodes, 500 kW klystrons, 500 kW gyrotrons, very low temperature (3 deg. K) electromechanical pumps, etc.. Tore Supra is a good example of the various applications of electricity and a testimony of the constant progress of the techniques mastered by electricians. This chapter is divided in 5 parts. Part 1 gives some general informations about thermonuclear fusion research, tokamak principles and electrotechnical systems of fusion research devices. Part 2 describes the Tore Supra tokamak, its aims and specificities, its internal components, the poloidal field system and the plasma heating systems. Part 3 concerns the power pulse sources: distribution network, poloidal field power supply, plasma heating systems, and ergodic divertor power supply. Part 4 describes the permanent electric power supplies for the auxiliary systems: toroidal field, cryogenic installation, cooling-drying loops. The last chapter briefly summarizes the perspectives of nuclear fusion research. (J.S.)

  20. Termoska pro tokamak

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2014-01-01

    Roč. 7, prosinec (2014), s. 16-17 Institutional support: RVO:61389021 Keywords : fusion * tokamak * cryostat * ITER Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1604-termoska-pro-tokamak

  1. Edge plasma diagnostics in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Brotánková, Jana; Hron, Martin; Adámek, Jiří; Ďuran, Ivan; Van Oost, G.; Peleman, P.; Gunn, J.; Devynck, P.; Martines, E.; Schrittwieser, R.; Kocan, M.

    Kudowa Zdrój : -, 2006, s. 910-935. [Sixth International Workshop and Summer School Towards Fusion Energy - Plasma Physics, Diagnostics, Spin-offs. Kudowa Zdrój (PL), 18.09.2006-22.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * diagnostics * heating Subject RIV: BL - Plasma and Gas Discharge Physics

  2. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  3. The impact of improved physics on commercial tokamak reactors

    International Nuclear Information System (INIS)

    Improvements in the confinement and beta capability of tokamak devices have long been a goal of the fusion program. We examine the impact of improvements in present day confinement and beta capabilities on commercial tokamak reactors. We characterize confinement with the achievable enhancement factor (H) over the ITER89 Power scaling confinement time, and beta by the Troyon coefficient g. A surprisingly narrow range of plasma confinement and beta are found to be useful in minimizing the cost of electricity for a tokamak reactor. Improvements in only one of these quantities is not useful beyond some point, without accompanying improvements in the other. For the plasma beta limited by a Troyon coefficient (g) near 4.3 (%mT/MA), confinement levels characterized by H factor enhancements of only 2 are useful for our nominal steady-state driven tokamak. These confinement levels are similar to those observed in present day experiments. If the permissible Troyon beta coefficient is near 6, the useful H factor confinement range increases to 2.5, still close to present day confinement levels. Inductively driven, pulsed reactors have somewhat increased useful ranges of confinement, relative to the steady-state cases. For a Troyon beta limit coefficient g near 4.3, H factors up to 2.5 are useful, and for g near 6, H factors up to 3 are useful

  4. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

    2002-07-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  5. Tokamak Systems Code

    International Nuclear Information System (INIS)

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

  6. Control of a burning tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Burmeister, R.E.; Mandrekas, J.; Stacey, W.M.

    1993-03-01

    This report is a review of the literature relevant to the control of the thermonuclear burn in a tokamak plasma. Some basic tokamak phenomena are reviewed, and then control by modulation of auxiliary heating and fueling is discussed. Other possible control methods such as magnetic ripple, plasma compression, and impurity injection as well as more recent proposed methods such as divertor biasing and L- to H-mode transition are also reviewed. The applications of modern control theory to the tokamak burn control problem are presented. The control results are summarized and areas of further research are identified.

  7. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

  8. Health physics around a controlled fusion research device: the Tokamak at Fontenay-aux-Roses (T.F.R.)

    International Nuclear Information System (INIS)

    The X and neutron dosimetry measurement near the magnetic confinement device for hot plasma, called T.F.R. (Tokamak, Fontenay-aux-Roses) are presented. The biological shielding consists of an ordinary concrete wall 30 cm thick; the dose rate is thus limited at 10-1 mrem per discharge (corresponding to 10 mrem per day) in the whole area frequented by people during T.F.R. operation. A numerical calculation, taking into account the true geometry and X ray reflexion by the walls and roof, and normalized to the measurements, gives some indications on the electron beam which produces X rays. The photoneutron source (up to 1010 neutrons per dischage) and the activation of the vacuum vessel result from high energy electrons (>= 10 MeV) supporting a 10 to 1,000 A current

  9. Recent Progress on Spherical Torus Research

    Energy Technology Data Exchange (ETDEWEB)

    Ono, Masayuki [PPPL; Kaita, Robert [PPPL

    2014-01-01

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ~ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ~ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of attractive fusion energy power source. Since the start of the two megaampere class ST facilities in 2000, National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all of fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

  10. Research and Development Supporting a Next Generation Germanium Double Beta Decay Experiment

    Science.gov (United States)

    Rielage, Keith; Elliott, Steve; Chu, Pinghan; Goett, Johnny; Massarczyk, Ralph; Xu, Wenqin

    2015-10-01

    To improve the search for neutrinoless double beta decay, the next-generation experiments will increase in source mass and continue to reduce backgrounds in the region of interest. A promising technology for the next generation experiment is large arrays of Germanium p-type point contact detectors enriched in 76-Ge. The experience, expertise and lessons learned from the MAJORANA DEMONSTRATOR and GERDA experiments naturally lead to a number of research and development activities that will be useful in guiding a future experiment utilizing Germanium. We will discuss some R&D activities including a hybrid cryostat design, background reduction in cabling, connectors and electronics, and modifications to reduce assembly time. We acknowledge the support of the U.S. Department of Energy through the LANL/LDRD Program.

  11. ITER tokamak device

    Science.gov (United States)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER, a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fueling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  12. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    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

  13. Edge plasma studies on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Hron, Martin; Peleman, P.; Spolaore, M.; Martines, E.; Hronová-Bilyková, Olena; Dejarnac, Renaud; Devynck, P.; Brotánková, Jana; Sentkerestiová, Jana; Ďuran, Ivan; Gunn, J.; Stöckel, Jan; Van Oost, G.; Adámek, Jiří; van de Peppel, L.; Štěpán, Michal

    Krakow : Euratom - IPPLM Association, 2006 - (Zagorski, R.), - [IEA Large Tokamak IA Workshop on Edge Transport in Fusion plasmas. Kraków (PL), 11.09.2006-13.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * scrape-off layer * turbulence * interchange instability Subject RIV: BL - Plasma and Gas Discharge Physics http://www.etfp2006.ifpilm.waw.pl/presentations.html

  14. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    The DIII-D tokamak research program is carried out by General Atomics for the U.S. Department of Energy. The DIII-D is the most flexible and best diagnosed tokamak in the world and the second largest tokamak in the U.S. The primary goal of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated in three major areas: Tokamak Physics, Divertor and Boundary Physics, and Advanced Tokamak Studies

  15. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  16. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  17. Tokamak concept innovations

    International Nuclear Information System (INIS)

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  18. Confinement of a self-stabilized tokamak under average magnetic well conditions

    International Nuclear Information System (INIS)

    It is well known that the average favorable magnetic curvature of a tokamak is stabilizing with respect to pressure-driven magnetohydrodynamic instabilities at low beta and that self-stabilization occurs at finite beta in the so-called second stability regime. Here we self-consistently investigate how these two effects, viz., the mean magnetic well and the self-stabilization, influence the energy confinement time in a tokamak, using the ballooning mode transport model

  19. Summary Report of 1st Research Coordination Meeting on Development of Reference Database for Beta-delayed Neutron Emission

    International Nuclear Information System (INIS)

    A summary is given of the 1st Research Coordination Meeting of the new IAEA Coordinated Research Project (CRP) on Development of a Reference Database for Beta-delayed neutron emission data. Participants presented their work, reviewed the current status of the field with regards to individual precursors and aggregate data, and discussed the scope of the work to be undertaken. A list of priorities and task assignments was produced. (author)

  20. 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

  1. Beta-limiting Instabilities and Global Mode Stabilization in NSTX

    Science.gov (United States)

    Sabbagh, Steven

    2001-10-01

    Low aspect ratio and high edge q theoretically alter the plasma stability and mode structure compared to standard tokamak configurations. Below the no-wall limit, stability calculations with PEST, GATO, and DCON show the perturbed radial field is maximized near the center column and DCON and VALEN calculations show that mode stability is not greatly improved by a nearby conducting wall due to the short poloidal wavelength in this region. In contrast, as beta reaches and exceeds the no-wall limit, the mode becomes strongly ballooning with long poloidal wavelength at large major radius and is highly wall stabilized. In this way, wall stabilization is more effective at higher beta in low aspect ratio geometry. Research on the stability of spherical torus plasmas at and above the no-wall beta limit is being addressed on NSTX, which has produced low aspect ratio plasmas, R/a = 1.27 at plasma current up to 1.4 MA with high energy confinement (TauE/TauE-ITER89P = 2). Toroidal and normalized beta have reached 22%, and 4.3, respectively in q = 7 plasmas. The beta limit is observed to increase with increasing plasma internal inductance, li, and the stability factor betaN/li has reached 5.8, limited by sudden beta collapses at low li that was achieved by use of high-harmonic fast wave heating (HHFW). DCON stability analysis of equilibria reconstructed with EFIT using external magnetics show that the plasmas are below or at the no-wall beta limit for the n = 1 mode, which has characteristics of a current-driven kink. With more peaked current profiles (li greater than 0.7), core MHD instabilities are observed which saturate or slowly degrade beta. Sawteeth with large inversion radii can also cause substantial beta collapses, although current profile modification using HHFW, altered plasma growth, and increased toroidal field have each been successful in mitigating this effect.

  2. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance

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

    International Nuclear Information System (INIS)

    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

  4. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    HUMPHREYS,D.A; FERRON,J.R; JOHNSON,R.D; LEUER,J.A; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M

    2003-10-01

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance.

  5. Design and construction of electronic components for a ''Novillo'' Tokamak

    International Nuclear Information System (INIS)

    The goal of this effort was to design, construct and make functional the electronic components for a ''Novillo'' Tokamak currently being experimentally investigated at the National Institute of Nuclear Research in Mexico. The problem was to develop programmable electronic switches capable of discharging high voltage kilowatt energies stored in capacitator banks onto the coils of the Tokamak. (author)

  6. Progress and prospects in understanding the physics of tokamak experiments

    International Nuclear Information System (INIS)

    A whistle-stop tour of the diverse physics of tokamak plasma confinement. This talk will illustrate the way in which fusion research on tokamaks has led to important and interesting physics results, and discuss some of the scientific challenges still ahead before fusion's potential can be established

  7. Role of the tokamak ISTTOK on the EURATOM fusion programme

    International Nuclear Information System (INIS)

    This paper describes the role of the tokamak ISTTOK on the development of the portuguese fusion research team, in the frame of the EURATOM Fusion Programme. Main tasks on education and training, control and data acquisition, diagnostics and tokamak physics are summarized. Work carried out on ISTTOK in collaboration with foreign teams is also reported. (author)

  8. Measurement and analysis of the radiation losses in DAMAVAND Tokamak

    International Nuclear Information System (INIS)

    Radiation losses play an important role on reaching to break-even conditions in Tokamaks. In this paper the results of measurement by a bolometer in Damavand Tokamak have been presented and analyzed. Meanwhile, we have explained our future research program on the base of last modifications in the control system of the DAMAVAND.

  9. Measurement of helicon wave coupling for current drive and anticipated role for high beta KSTAR plasmas

    Science.gov (United States)

    Wang, S. J.; Kim, H. J.; Joung, M.; Jeong, J. H.; Kim, J. H.; Bae, Y. S.; Kwak, J. G.; Wi, H. H.; Kim, H.-S.

    2015-11-01

    Helicon wave current drive has been suggested for efficient off-axis current drive in high electron beta tokamak plasmas. Fast wave drives centrally peaking current in the frequency range up to several ion cyclotron harmonics in the present tokamaks, such as KSTAR. Increasing fast wave frequency up to LH resonance frequency at the plasma edge, the spiral propagation of wave at the outer region of plasma lengthens the wave path to the plasma center. Also, optical thickness increases with frequency. It is expected that these effects produce efficient off-axis power deposition depending on the electron beta and magnetic field pitch. A low power TWA for helicon wave was installed and tested in KSTAR tokamak which is aiming for the steady-state high beta plasma requiring off-axis current drive. The power coupling properties of TWA at various plasma conditions will be presented. In addition to the coupling efficiency, issues such as load sensitivity and unwanted slow wave coupling will be addressed. Also, the simulation of plasma performance with the combination of helicon wave current drive and other conventional heating and current drive power in KSTAR will be discussed. This work was supported by the Korean Ministry of Science, ICT and Future Planning and by R&D Program through the National Fusion Research Institute of Korea (NFRI) funded by the Government funds.

  10. Tokamak ARC damage

    International Nuclear Information System (INIS)

    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

  11. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  12. Microinstability theory in tokamaks: a review

    International Nuclear Information System (INIS)

    Significant investigations in the area of tokamak microinstability theory are reviewed. Emphasis is given to the work covering the period from 1970 through 1976. Special attention is focused on low-frequency electrostatic drift-type modes, which are generally believed to be the dominant tokamak microinstabilities under normal operating conditions. The basic linear formalism including electromagnetic (finite beta) modifications is presented along with a general survey of the numerous papers investigating specific linear and nonlinear effects on these modes. Estimates of the associated anomalous transport and confinement times are discussed, and a summary of relevant experimental results is given. Studies of the nonelectrostatic and high-frequency instabilities associated with the presence of high energy ions from neutral beam injection (or with the presence of alpha particles from fusion reactions) are also surveyed

  13. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír

    2012-01-01

    Roč. 87, č. 12 (2012), s. 2238-2241. ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006

  14. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    The DIII-D tokamak research program is carried out by, General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data needed by International Thermonuclear Experimental Reactor (ITER) and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics

  15. The Thor tokamak experiment

    International Nuclear Information System (INIS)

    The main characteristics of the plasma produced in Thor tokamak discharges are described. The machine performances are outlined and the experimental results relevant to the equilibrium, the stability and the control of the discharge regimes are discussed in detail. (author)

  16. Modular tokamak magnetic system

    Science.gov (United States)

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  17. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-01-01

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

  18. Tokamak simulation code manual

    International Nuclear Information System (INIS)

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

  19. US/French joint research program regarding the behavior of polymer base materials subjected to beta radiation

    International Nuclear Information System (INIS)

    This document is the final account of a multiyear joint NRC/CEA cooperative research program to investigate the relative effectiveness of beta and gamma irradiations to produce in polymer base electrical insulation and jacket materials. During the program jointly executed by research laboratories in the US (Sandia National Laboratories) and France (Compagnie ORIS Industrie), a number of material and radiation parameters were investigated. Results obtained were reasonably independent of the radiation parameters and most material parameters investigated and would suggest that material damage resulting from electron beam and gamma ray irradiations can be correlated, within limits, on the basis of absorbed radiation dose. 7 refs., 26 figs., 13 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J. [General Atomics, San Diego, CA (United States); Lazarus, E. [Oak Ridge National Lab., TN (United States)

    1991-11-01

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

  1. The Experiments of the small Spherical Tokamak Gutta

    International Nuclear Information System (INIS)

    GUTTA is a small spherical tokamak (R = 16cm, a = 8cm, Ip = 150kA) operating at the St. Petersburg State University since 2004 in the scope of the IAEA CRP ''Joint Research using Small Tokamaks''. Main scientific activities on GUTTA include development of new and improvement of existing mathematical models of plasma control, relevant for application on large tokamaks and ITER and verification of them on GUTTA; studies on the ECRH/EBW assisted breakdown and non-solenoid plasma formation in low aspect ratio tokamak; development of diagnostics; training and education of students.In this paper design properties of Gutta will be presented. Regimes of operation of the tokamak and plasma shape parameters are described and first results of the plasma formation and start-up studied will be discussed

  2. Recent experiments and upgradation plans for Aditya Tokamak

    International Nuclear Information System (INIS)

    Several experiments relevant to the operation of future big tokamaks including ITER and also contributing significantly to the tokamak based thermonuclear fusion research have been carried out in Aditya tokamak recently. Low loop voltage start-up of plasma current has been successfully obtained with ICR and ECR preionization. Reduced runaway generation is achieved by applying a local vertical magnetic field at one toroidal location. Plasma disruptions, a sudden loss of equilibrium and confinement, has been successfully mitigated through application of bias voltage on a Molybdenum (Mo) electrode placed inside the last closed flux surface (LCFS). Extensive studies on plasma flows, effect of gas puff on flows in the Scrape off layer (SOL) and impurity transport has been carried out. Effect of Helium glow discharge cleaning (GDC) on partial pressures and plasma parameters have also been studied for plasma performance improvement. To contribute more to the bigger tokamaks operated in the divertor configuration, the existing Aditya tokamak with limiter configuration, which is in operation for 24 years, is planned to be upgraded to a divertor machine. The main aim of the Aditya-U tokamak is to have a small/mid-size tokamak with divertor operation and higher duty cycle, which will be test bed for new diagnostics, the students can be trained and those experiments can be tried out which are not desirable in big tokamaks, such as runaway mitigation and disruption control. Details of experimental results and upgradation plan will be discussed in the talk. (author)

  3. Electron cyclotron emission diagnostics on KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, S. H. [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Daejeon 305-353 (Korea, Republic of); Lee, K. D.; Kwon, M. [National Fusion Research Institute, 113 Gwahangno, Daejeon 305-333 (Korea, Republic of); Kogi, Y. [Fukuoka Institute of Technology, Higashiku, Fukuoka 811-0295 (Japan); Kawahata, K.; Nagayama, Y. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Mase, A. [KASTEC, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  4. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration. PMID:21033954

  5. Stability of Tokamaks with respect to slip motions

    International Nuclear Information System (INIS)

    Using the energy principle in Tokamaks we investigate a class of perturbations which, if unstable, cannot be stabilized by the toroidal main field. On the assumptions of usual Tokamak ordering and in the limit of infinite aspect ratio, these perturbations are shown to be minimizing among all axisymmetric perturbations. In the case of finite aspect ratio, a detailed stability analysis is carried out using a constant pressure surface current model with elliptic, triangular or rectangular plasma cross-section. Definite stabilization by toroidal effects and by beta poloidal is demonstrated. (orig.)

  6. Research, development and application of noncombustible Beta fiber structures. [for Apollo

    Science.gov (United States)

    Dillon, J. J.; Cobb, E. S.

    1975-01-01

    Beta fiber was selected as the primary material for flexible fibrous structures used in spacecraft and crew systems applications in the Apollo program because it was noncombustible in a 100 percent oxygen atmosphere up to 16.5 psia. It met NASA criteria for outgassing, toxicity, odor, and crew comfort, and possessed sufficient durability to last through the mission. Topics discussed include: study of spacecraft applications; design of Beta fiber textile structures to meet the requirements; selection of surface treatments (finishes, coatings, and printing systems) to impart the required durability and special functional use to the textile structures; development of sewing and fabrication techniques; and testing and evaluation programs, and development of production sources.

  7. Target plasma achievements for unity beta operation in ET

    Science.gov (United States)

    Gourdain, P.-A.; Carter, T. A.; Gauvreau, J.-L.; Grossman, A.; Lafonteese, D. J.; Pace, D. C.; Schmitz, L. W.; Taylor, R. J.

    2003-10-01

    The Electric Tokamak (ET) at UCLA (t_pulse0.3V, B=0.25T) is now in a critical stage of research to exceed the conventional beta limit. The latest program is aiming for 10s long plasma discharges at unity peak beta by means of 2^d harmonic ICRF heating and current profile control. A stable computational MHD high beta path has been found for ET high aspect ratio configuration leading from the Troyon limit to unity beta using an in-house equilibrium code and the DCON stability code. This critical path involves high power ICRF injection (5MW) for targeted plasma heating and ballooning/kink modes stabilization by poloidal rotation. Another major component is high plasma density using core and edge fueling relying on kinetic neutral beam and inboard/outboard gas puff. At this regime current profile control using fast wave current drive, with bootstrap current optimization, will be possible. A magnetic well should develop, leading to high beta plasmas with omnigenous properties. To achieve these goals, a tight feedback system between power and particle injections and plasma diagnostics is mandatory.

  8. Tokamak experimental section

    International Nuclear Information System (INIS)

    Descriptions of research during this period are given for the following topics: (1) ion and electron heating, (2) high-beta and gas puff experiments, (3) beam trapping by impurities, (4) counterinjection studies, (5) impurity measurements, (6) Balmer alpha line profiles, (7) internal mode structure, (8) sawtooth oscillations and plasma transport, (9) Ormak plasma modeling, (10) charge exchange measurements, (11) wall power measurements, (12) neutron time behavior due to deuterium neutral beam injection into a hydrogen plasma, (13) wall impurities in Ormak, (14) relativistic electron studies, (15) fast x-ray energy analyzer for the 1 to 10 keV range, and (16) CTR related atomic physics

  9. Natural elongation and triangularity of tokamak equilibria

    International Nuclear Information System (INIS)

    Quasianalytic formulas are calculated for the elongation κ and triangularity δ of the plasma surface of a free-boundary tokamak equilibrium. The final results give κ and δ as functions of five quantities: the inverse aspect ratio ε, the poloidal beta βp, the internal inductance li, and the quadrupole and hexapole moments of the externally applied field. The agreement with numerically computed equilibria is found to be quite good when A≥3, κ≤1.5, and δ≤0.2 and when the plasma is limited by the vacuum vessel wall and not diverted by the presence of a separatrix on the plasma surface

  10. DIII-D research operations annual report to the U.S. Department of Energy, October 1, 1995--September 30, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The mission of the DIII-D research program is to advance fusion energy science understanding and predictive capability and to improve and optimize the tokamak concept. A long term goal remains to integrate these products into a demonstration of high confinement, high plasma pressure (plasma {beta}), sustained long pulse operation with fusion power plant relevant heat and particle handling capability. The DIII-D program is a world recognized leader in tokamak concept improvement and a major contributor to the physics R and D needs of the International Thermonuclear Experimental Reactor (ITER). The scientific objectives of the DIII-D program are given in Table 1-2. The FY96 DIII-D research program was highly successful, as described in this report. A moderate sized tokamak, DIII-D is a world leader in tokamak innovation with exceptional performance, measured in normalized parameters.

  11. Vesmírný tokamak na Zemi

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2007-01-01

    Roč. 15, č. 3 (2007), s. 12-14. ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * technology * material * tokamak * ITER Subject RIV: BL - Plasma and Gas Discharge Physics

  12. Design of selected subsystems for COMPASS tokamak operation

    Czech Academy of Sciences Publication Activity Database

    Janky, Filip; Pereira, T.; Hron, Martin; Pánek, Radomír; Fernandes, H.

    Aix-en-Provence : IAEA, 2009. s. 80-80. ISBN N. [Seventh IAEATechnical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research. 15.06.2009-19.06.2009, Aix-en-Provence] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * Compass * machine control * tokamak operation Subject RIV: BL - Plasma and Gas Discharge Physics http://www-fusion-magnetique.cea.fr/tmiaea2009/ website /data/articles/000080.pdf

  13. A nonlinear magnetic helicity model of a tight aspect ratio bootstrapped tokamak with oscillating field current drive

    International Nuclear Information System (INIS)

    Theoretical and computational analyses of bootstrapped tokamaks with oscillating field current drive (OFCD) have been developed using a nonlinear magnetic helicity model. Assuming a rigid current profile, the conditions for optimal tokamak steady-state operation are derived and are shown to agree with the results of computer calculations. Generalized limit formulae for the toroidal plasma beta and bootstrap current fraction are also obtained. The results of the analyses indicate that steady-state bootstrapped tokamak operation with high toroidal plasma beta can best be achieved in tight aspect ratio A → 1 reactors with OFCD supplying as much as a quarter of the total toroidal plasma current. (author)

  14. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

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

  15. Research of TGF-beta1 Inducing Lung Adencarcinoma PC9 Cells to Mesenchymal Cells Transition

    OpenAIRE

    Chen, Xiaofeng; Wang, Heyong; Zhang, Lei; Zhang, Huijun

    2010-01-01

    Background and objective It has been proven that epithelial-mesenchymal transition (EMT) not only correlated with embryonic development but also could promote tumor invasion and metastasis. Transforming growth factor beta-1 (TGF-β1) has been identified as the main inducer of tumor EMT. The aim of this study was to investigate the effects of TGF-β1 on EMT and PI3K/AKT signaling pathway in lung adencarcinoma PC9 cells. Methods Cultured PC9 cells were treated with different concentrations of TGF...

  16. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    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

  17. The National Spherical Torus Experiment (NSTX) research program and progress towards high beta, long pulse operating scenarios

    International Nuclear Information System (INIS)

    A major research goal of the National Spherical Torus Experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed over the last two years, including neutral beam (up to 7 MW) and high harmonic fast wave heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with T> up to 35%. Normalized beta values often exceed the no wall limit, and studies suggest that passive wall mode stabilization is enabling this for broad pressure profiles characteristic of H mode plasmas. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fueling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is by comparing of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. A peak heat flux of 10 MW/m2 has been measured in the H mode, with large asymmetries in the power deposition being observed between the inner and outer strike points. Noninductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun. (author)

  18. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    International Nuclear Information System (INIS)

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

  19. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    Science.gov (United States)

    Azizov, E. A.

    2012-02-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.

  20. Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

    International Nuclear Information System (INIS)

    A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the 'advanced tokamak' physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs

  1. Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)

    1994-11-01

    A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.

  2. [Fusion energy research

    International Nuclear Information System (INIS)

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer

  3. (Fusion energy research)

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, C.A. (ed.)

    1988-01-01

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer.

  4. UCLA Tokamak Program Close Out Report.

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Robert John [UCLA/retired

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  5. Material erosion and migration in tokamaks

    International Nuclear Information System (INIS)

    The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the subsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasma-facing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined

  6. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Full text: SST-1, a steady state superconducting tokamak, is at advanced stage of erection at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation and triangularity. The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 T at plasma center and a plasma current of 220 kA. Hydrogen gas will be used and plasma discharge duration will be 1000 s. 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, made up of 16 vessel sectors having ports and 16 rings with D- shaped cross-section, which are welded in-situ during the SST-1 assembly. 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. In SST-1 tokamak, the auxiliary current drive will be 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. The assembly of the SST-1 tokamak is nearing completion. The cool down of the Superconducting magnets is scheduled to start by middle of year 2004

  7. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  8. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1996-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  9. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1998-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  10. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  11. IFS Numerical Laboratory Tokamak

    International Nuclear Information System (INIS)

    A numerical laboratory of a tokamak plasma is being developed. This consists of the backbone (the overall manager in terms of the MPPL programming language), and the modularized components that can be plugged in or out for a particular run and their hierarchical arrangement. The components include various metrics for overall geometry various dynamics, field calculations, and diagnoses. 2 refs

  12. Recent progress on spherical torus research

    International Nuclear Information System (INIS)

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ∼ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ∼ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment in the United States and Mega Ampere Spherical Tokamak in UK, active ST research has been conducted worldwide. More than 16 ST research facilities operating during this period have achieved remarkable advances in all fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era

  13. Beta Thalassemia

    Science.gov (United States)

    ... South Asian (Indian, Pakistani, etc.), Southeast Asian and Chinese descent. 1 Beta Thalassemia ßß Normal beta globin ... then there is a 25% chance with each pregnancy that their child will inherit two abnormal beta ...

  14. Transport in gyrokinetic tokamaks

    International Nuclear Information System (INIS)

    A comprehensive study of transport in full-volume gyrokinetic (gk) simulations of ion temperature gradient driven turbulence in core tokamak plasmas is presented. Though this ''gyrokinetic tokamak'' is much simpler than experimental tokamaks, such simplicity is an asset, because a dependable nonlinear transport theory for such systems should be more attainable. Toward this end, we pursue two related lines of inquiry. (1) We study the scalings of gk tokamaks with respect to important system parameters. In contrast to real machines, the scalings of larger gk systems (a/ρs approx-gt 64) with minor radius, with current, and with a/ρs are roughly consistent with the approximate theoretical expectations for electrostatic turbulent transport which exist as yet. Smaller systems manifest quite different scalings, which aids in interpreting differing mass-scaling results in other work. (2) With the goal of developing a first-principles theory of gk transport, we use the gk data to infer the underlying transport physics. The data indicate that, of the many modes k present in the simulation, only a modest number (Nk ∼ 10) of k dominate the transport, and for each, only a handful (Np ∼ 5) of couplings to other modes p appear to be significant, implying that the essential transport physics may be described by a far simpler system than would have been expected on the basis of earlier nonlinear theory alone. Part of this analysis is the inference of the coupling coefficients Mkpq governing the nonlinear mode interactions, whose measurement from tokamak simulation data is presented here for the first time

  15. Recent experiments on the STOR-M Tokamak

    International Nuclear Information System (INIS)

    Recent experiments on the STOR-M tokamak have been focused on basic tokamak physics and technology development for controlled thermonuclear fusion research. Active control of the magnetohydrodynamic (MHD) instabilities has been achieved by helical resonant magnetic perturbations (RMPs). Improved confinement has been induced by gas puffing during ohmic discharges. Modification of toroidal flow velocities by a tangentially injected compact torus (CT) plasmoid to the STOR-M discharge has been observed. (author)

  16. The Texas Experimental Tokamak: A plasma research facility. A proposal submitted to the Department of Energy in response to Program Notice 95-10: Innovations in toroidal magnetic confinement systems

    International Nuclear Information System (INIS)

    The Fusion Research Center (FRC) at the University Texas will operate the tokamak TEXT-U and its associated systems for experimental research in basic plasma physics. While the tokamak is not innovative, the research program, diagnostics and planned experiments are. The fusion community will reap the benefits of the success in completing the upgrades (auxiliary heating, divertor, diagnostics, wall conditioning), developing diverted discharges in both double and single null configurations, exploring improved confinement regimes including a limiter H-mode, and developing unique, critical turbulence diagnostics. With these new regimes, the authors are poised to perform the sort of turbulence and transport studies for which the TEXT group has distinguished itself and for which the upgrade was intended. TEXT-U is also a facility for collaborators to perform innovative experiments and develop diagnostics before transferring them to larger machines. The general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The emphasis of the proposed research is to provide well-documented plasmas which will be used to suggest and evaluate theories, to explore control techniques, to develop advanced diagnostics and analysis techniques, and to extend current drive techniques. Up to 1 MW of electron cyclotron heating (ECH) will be used not only for heating but as a localized, perturbative tool. Areas of proposed research are: (1) core turbulence and transport; (2) edge turbulence and transport; (3) turbulence analysis; (4) improved confinement; (5) ECH physics; (6) Alfven wave current drive; and (7) diagnostic development

  17. Spontaneous generation of rotation in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Parra Diaz, Felix [Oxford University

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  18. Radioactivity evaluation for the KSTAR tokamak

    International Nuclear Information System (INIS)

    The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 1016 s-1 through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10-4 mrem h-1 in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 106 Bq kg-1 in the carbon graphite of a plasma-facing wall. (authors)

  19. Module description of TOKAMAK equilibrium code MEUDAS

    International Nuclear Information System (INIS)

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  20. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-01-01

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  1. Draft program plant for TNS: The Next Step after the tokamak fusion test reactor. Part III. Project specific RD and D needs

    International Nuclear Information System (INIS)

    Research and development needs for the TNS systems are described according to the following chapters: (1) tokamak system, (2) electrical power systems, (3) plasma heating systems, (4) tokamak support systems, (5) instrumentation, control, and data systems, and (6) program recommendations

  2. Research of TGF-beta1 Inducing Lung Adencarcinoma PC9 Cells to Mesenchymal Cells Transition

    Directory of Open Access Journals (Sweden)

    Xiaofeng CHEN

    2010-01-01

    Full Text Available Background and objective It has been proven that epithelial-mesenchymal transition (EMT not only correlated with embryonic development but also could promote tumor invasion and metastasis. Transforming growth factor beta-1 (TGF-β1 has been identified as the main inducer of tumor EMT. The aim of this study was to investigate the effects of TGF-β1 on EMT and PI3K/AKT signaling pathway in lung adencarcinoma PC9 cells. Methods Cultured PC9 cells were treated with different concentrations of TGF-β1 for 48 h. The morphological changes were observed under phase-contrast microscopy; EMT relative marker protein changes were assessed by Western blot and immunoflurescence staining. In addition, the expression of AKT and P-AKT were also measured by Western blot. Results The data showed that TGF-β1 could induce PC9 morphological alteration from epithelial to mesenchymal and upregulate the expression of mesenchymal maker protein Fibronectin. Obviously, the expression of P-AKT was downregulated by TGF-β1 treatment for 48 h. Conclusion TGF-β1 might induce EMT of PC9 cells , accompanied by the changes of PI3K/AKT signaling pathway.

  3. DIII-D research operations

    International Nuclear Information System (INIS)

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R ampersand D; and collaborative efforts

  4. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    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

  5. A wide variety of putative extremophiles and large beta-diversity at the Mars Desert Research Station (Utah)

    Science.gov (United States)

    Direito, Susana O. L.; Ehrenfreund, Pascale; Marees, Andries; Staats, Martijn; Foing, Bernard; Röling, Wilfred F. M.

    2011-07-01

    Humankind's innate curiosity makes us wonder whether life is or was present on other planetary bodies such as Mars. The EuroGeoMars 2009 campaign was organized at the Mars Desert Research Station (MDRS) to perform multidisciplinary astrobiology research. MDRS in southeast Utah is situated in a cold arid desert with mineralogy and erosion processes comparable to those on Mars. Insight into the microbial community composition of this terrestrial Mars analogue provides essential information for the search for life on Mars: including sampling and life detection methodology optimization and what kind of organisms to expect. Soil samples were collected from different locations. Culture-independent molecular analyses directed at ribosomal RNA genes revealed the presence of all three domains of life (Archaea, Bacteria and Eukarya), but these were not detected in all samples. Spiking experiments revealed that this appears to relate to low DNA recovery, due to adsorption or degradation. Bacteria were most frequently detected and showed high alpha- and beta-diversity. Members of the Actinobacteria, Proteobacteria, Bacteroidetes and Gemmatimonadetes phyla were found in the majority of samples. Archaea alpha- and beta-diversity was very low. For Eukarya, a diverse range of organisms was identified, such as fungi, green algae and several phyla of Protozoa. Phylogenetic analysis revealed an extraordinary variety of putative extremophiles, mainly Bacteria but also Archaea and Eukarya. These comprised radioresistant, endolithic, chasmolithic, xerophilic, hypolithic, thermophilic, thermoacidophilic, psychrophilic, halophilic, haloalkaliphilic and alkaliphilic micro-organisms. Overall, our data revealed large difference in occurrence and diversity over short distances, indicating the need for high-sampling frequency at similar sites. DNA extraction methods need to be optimized to improve extraction efficiencies.

  6. Next tokamak facility

    International Nuclear Information System (INIS)

    Design studies on a superconducting, long-pulse, current-driven, ignited tokamak, called the Toroidal Fusion Core Demonstration (TFCD), are being conducted by the Fusion Engineering Design Center (FEDC) and Princeton Plasma Physics Laboratory (PPPL) with additional broad community involvement. Options include the use of all-superconducting toroidal field (TF) coils, a superconducting-copper hybrid arrangement of TF coils, or all-copper TF coils. Only the first two options have been considered to date. The general feasibility of these approaches has been established with the goal of high performance (ignition, approx. 390 MW; wall loading approx. 2.2 MW/m2) at minimum capital cost. The preconceptual effort will be completed in early FY 1984 and a selection made from the indicated options. The TFCD is judged to represent a reasonable necessary step between the Tokamak Fusion Test Reactor (TFTR) and the Engineering Test Reactor

  7. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  8. Tritium catalyzed deuterium tokamaks

    International Nuclear Information System (INIS)

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the 3He from the D(D,n)3He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general)

  9. Tokamak pump limiters

    International Nuclear Information System (INIS)

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

  10. Microwave correllation reflectometry for tokamak CASTOR

    Czech Academy of Sciences Publication Activity Database

    Nanobashvili, S.; Žáček, František; Zajac, Jaromír

    2005-01-01

    Roč. 55, č. 6 (2005), s. 701-719. ISSN 0011-4626 R&D Projects: GA AV ČR IAA1043101 Grant ostatní: GA EU(EU) INTAS ´2001 1B-2056 Institutional research plan: CEZ:AV0Z20430508 Keywords : microwaves * tokamak * plasma * turbulence * reflectometry Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.360, year: 2005

  11. Frascati Tokamak Upgrade (FTU): Results and developments

    International Nuclear Information System (INIS)

    In the present note the relation is examined between the FTU experimental programme and the most important issues in controlled thermonuclear fusion researches. FTU is a high-density, high magnetic field tokamak devoted to the study of plasma heating and current drive, energy and particle confinement and plasma-wall interaction. The most important FTU results and their relevance for ITER will be discussed

  12. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1993--September 30, 1994

    International Nuclear Information System (INIS)

    The DIII-D tokamak research program is managed by General Atomics (GA) for the US Department of Energy (DOE). Major program participants include GA, Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Laboratory (ORNL), and the University of California together with several other national laboratories and universities. The DIII-D is a moderate sized tokamak with great flexibility and extremely capable subsystems. The primary goal of the DIII-D tokamak research program is to provide data for development of a conceptual physics blueprint for a commercially attractive fusion power plant. In so doing, the DIII-D program provides physics and technology R ampersand D output to aid the International Thermonuclear Experimental Reactor (ITER) and the Princeton Tokamak Physics Experiment (TPX) projects. Specific DIII-D objectives include the achievement of steady-state plasma current as well as the demonstration of techniques for radio frequency heating, divertor heat removal, particle exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion in plasmas with high beta and with high confinement. The long-range plan is organized with two principal elements, the development of an advanced divertor and the development of advanced tokamak concepts. These two elements have a common goal: an improved demonstration reactor (DEMO) with lower cost and smaller size than present DEMO concepts. In order to prepare for this long-range development, in FY94 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak studies, and Tokamak Physics

  13. VDE characteristics during disruption process and its underlying acceleration mechanism in the ITER-EDA tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Yukiharu; Nishio, Satoshi; Yoshino, Ryuji [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kessel, C.E.; Jardin, S.C.

    1996-12-01

    The dynamic behavior of vertical displacement events (VDEs) during a disruption and acceleration mechanisms that govern VDEs in the ITER-EDA tokamak are investigated using the Tokamak Simulation Code. A sudden plasma pressure drop ({beta}{sub p} collapse) does not accelerate VDEs for the ITER tokamak. The geometry of the ITER resistive shell is shown to be suitable for preventing a {beta}{sub p} collapse-induced VDE, because the magnetic field decay n-index after the {beta}{sub p} collapse does not considerably degrade. On the other hand, it is shown that the plasma current quench (I{sub p} quench) following the energy quench can accelerate VDEs due to the vertical imbalance of the attractive force arising from the up-down asymmetric shell. The vertical location of the neutral point where the I{sub p} quench-induced VDE almost disappears is found to lie at {approx}22 cm below the plasma magnetic axis of the nominal equilibrium (Z = 1.44 m). An upward and moderate I{sub p} quench-induced VDE can be expected for the nominal configuration in the ITER-EDA tokamak. It is shown that the ITER tokamak has an advantage of avoiding the fatal damage of the complicated structures of the bottom-divertor. (author)

  14. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

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

  15. Polarization spectroscopy of tokamak plasmas

    International Nuclear Information System (INIS)

    Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs

  16. Physics issues of high bootstrap current tokamaks

    International Nuclear Information System (INIS)

    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 qmin is close to the region of high ∇ne. 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. Physics evaluation of compact tokamak ignition experiments

    International Nuclear Information System (INIS)

    At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/2/q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs

  18. DIII-D tokamak long range plan. Revision 3

    International Nuclear Information System (INIS)

    The DIII-D Tokamak Long Range Plan for controlled thermonuclear magnetic fusion research will be carried out with broad national and international participation. The plan covers: (1) operation of the DIII-D tokamak to conduct research experiments to address needs of the US Magnetic Fusion Program; (2) facility modifications to allow these new experiments to be conducted; and (3) collaborations with other laboratories to integrate DIII-D research into the national and international fusion programs. The period covered by this plan is 1 November 19983 through 31 October 1998

  19. Steady state operation of tokamaks. Report on the IAEA technical committee meeting held at Hefei, China, 13-15 October 1998

    International Nuclear Information System (INIS)

    The first IAEA Technical Committee Meeting on Steady State Operation of Tokamaks was held in October 1998 in Hefei, China. This meeting marks the timely start of Technical Committee Meetings in an important area of tokamak research since several experiments are already yielding impressive results and several new experiments are under construction. Among the ongoing experiments interesting results were reported from the superconducting tokamaks TRIAM 1-M, Tore Supra, and HT-7 and from a conventional tokamak, HL-1M

  20. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  1. Methods for the design and optimization of shaped tokamaks

    International Nuclear Information System (INIS)

    Two major questions associated with the design and optimization of shaped tokamaks are considered. How do physics and engineering constraints affect the design of shaped tokamaks? How can the process of designing shaped tokamaks be improved? The first question is addressed with the aid of a completely analytical procedure for optimizing the design of a resistive-magnet tokamak reactor. It is shown that physics constraints---particularly the MHD beta limits and the Murakami density limit---have an enormous, and sometimes, unexpected effect on the final design. The second question is addressed through the development of a series of computer models for calculating plasma equilibria, estimating poloidal field coil currents, and analyzing axisymmetric MHD stability in the presence of resistive conductors and feedback. The models offer potential advantages over conventional methods since they are characterized by extremely fast computer execution times, simplicity, and robustness. Furthermore, evidence is presented that suggests that very little loss of accuracy is required to achieve these desirable features. 94 refs., 66 figs., 14 tabs

  2. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  3. Understanding disruptions in tokamaks

    International Nuclear Information System (INIS)

    This paper describes progress achieved since 2007 in understanding disruptions in tokamaks, when the effect of plasma current sharing with the wall was introduced into theory. As a result, the toroidal asymmetry of the plasma current measurements during vertical disruption event (VDE) on the Joint European Torus was explained. A new kind of plasma equilibria and mode coupling was introduced into theory, which can explain the duration of the external kink 1/1 mode during VDE. The paper presents first results of numerical simulations using a free boundary plasma model, relevant to disruptions.

  4. Tokamak instrumentation and controls

    Energy Technology Data Exchange (ETDEWEB)

    Becraft, W. R.; Bettis, E. S.; Houlberg, W. A.; Onega, R. J.; Stone, R. S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine.

  5. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System.

  6. A Model Reference Adaptive Control for Radial Plasma Position on HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    MAOSuying; YUANBaoshan; LIQiang

    2003-01-01

    The radial plasma position control is one of the basic plasma controls of tokamak. In order to maintain a plasma column in the geometrical center of its vacuum vessel, the externally applied vertical field (VF) must be adjusted to the changes in the plasma parameters such as the plasma current, poloidal beta and the internal inductance.

  7. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  8. Beta-limiting instabilities and global mode stabilization in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Research on the stability of spherical torus plasmas at and above the no-wall beta limit is being addressed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)], that has produced low aspect ratio plasmas, R/a∼1.27 at plasma current exceeding 1.4 MA with high energy confinement (TauE/TauEITER89P>2). Toroidal and normalized beta have exceeded 25% and 4.3, respectively, in q∼7 plasmas. The beta limit is observed to increase and then saturate with increasing li. The stability factor βN/li has reached 6, limited by sudden beta collapses. Increased pressure peaking leads to a decrease in βN. Ideal stability analysis of equilibria reconstructed with EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] shows that the plasmas are at the no-wall beta limit for the n=1 kink/ballooning mode. Low aspect ratio and high edge q theoretically alter the plasma stability and mode structure compared to standard tokamak configurations. Below the no-wall limit, stability calculations show the perturbed radial field is maximized near the center column and mode stability is not highly effected by a nearby conducting wall due to the short poloidal wavelength in this region. In contrast, as beta reaches and exceeds the no-wall limit, the mode becomes strongly ballooning with long poloidal wavelength at large major radius and is highly wall stabilized. In this way, wall stabilization is more effective at higher beta in low aspect ratio geometry. The resistive wall mode has been observed in plasmas exceeding the ideal no-wall beta limit and leads to rapid toroidal rotation damping across the plasma core

  9. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  10. Relativistic runaway electrons in tokamak plasmas

    International Nuclear Information System (INIS)

    Runaway electrons are inherently present in a tokamak, in which an electric field is applied to drive a toroidal current. The experimental work is performed in the tokamak TEXTOR. Here runaway electrons can acquire energies of up to 30 MeV. The runaway electrons are studied by measuring their synchrotron radiation, which is emitted in the infrared wavelength range. The studies presented are unique in the sense that they are the first ones in tokamak research to employ this radiation. Hitherto, studies of runaway electrons revealed information about their loss in the edge of the discharge. The behaviour of confined runaways was still a terra incognita. The measurement of the synchrotron radiation allows a direct observation of the behaviour of runaway electrons in the hot core of the plasma. Information on the energy, the number and the momentum distribution of the runaway electrons is obtained. The production rate of the runaway electrons, their transport and the runaway interaction with plasma waves are studied. (orig./HP)

  11. Potential of commercial tokamak reactors with improved physics

    International Nuclear Information System (INIS)

    The impact of improvements in plasma confinement and beta on commercial tokamak embodiments has been systematically investigated. Cases were investigated for both steady-state and pulsed reactors. A narrow range of useful confinement enhancement (H factors over the ITER-89 Power scaling) and Troyon beta limit coefficients g are found to be beneficial for reducing the cost of electricity (COE). Improvements in the beta limit (i.e., g) offer the greatest improvement in reactors. For g limited to about 4, H factors of only 2 are needed. If g of 6 is attainable, H factors up to 2.5 are useful. Pulsed reactors have slightly higher useful H factors, ranging from 2.5 for g near 4 to 3 for g near 6

  12. Bifurcated Helical Core Equilibrium States in Tokamaks

    International Nuclear Information System (INIS)

    Full text: Tokamaks with weak to moderate reversed central magnetic shear in which the minimum of the inverse rotational transform qmin is in the neighbourhood of unity can trigger bifurcated MagnetoHydroDynamic (MHD) equilibrium states. In addition to the standard axisymmetric branch that can be obtained with standard Grad-Shafranov solvers, a novel branch with a three-dimensional (3D) helical core has been computed with the ANIMEC code, an anisotropic pressure extension of the VMEC code. The solutions have imposed nested magnetic flux surfaces and are similar to saturated ideal internal kink modes. The difference in energy between both possible branches is very small. Plasma elongation, current and β enhance the susceptibility for bifurcations to occur. An initial value nonlinear ideal MHD evolution of the axisymmetric branch compares favourably with the helical core equilibrium structures calculated. Peaked prescribed pressure profiles reproduce the 'snake' structures observed in many tokamaks which has led to a new explanation of the snake as a bifurcated helical equilibrium state that results from a saturated ideal internal kink in which pellets or impurities induce a hollow current profile. Snake equilibrium structures are computed in free boundary TCV tokamak simulations. Magnetic field ripple and resonant magnetic perturbations in MAST free boundary calculations do not alter the helical core deformation in a significant manner when qmin is near unity. These bifurcated solutions constitute a paradigm shift that motivates the application of tools developed for stellarator research in tokamak physics investigations. The examination of fast ion confinement in this class of equilibria is performed with the VENUS code in which a coordinate independent noncanonical phase-space Lagrangian formulation of guiding centre drift orbit theory has been implemented. (author)

  13. Tokamak exhaust process for the ITER project

    International Nuclear Information System (INIS)

    The ITER project calls for an unprecedented amount of hydrogen isotopes to be processed. To facilitate environmental responsibility and economic application of fusion technology, the re-use of hydrogen isotopes is vital. The US ITER Project Office (USIPO) is responsible for the front end of the ITER Tritium Plant, the Tokamak Exhaust Processing (TEP) System. The TEP system must separate the Tokamak exhaust gases into a stream containing only hydrogen isotopes and a stream containing only non-hydrogen gases. The USIPO has selected the Savannah River National Laboratory (SRNL) in partnership with the Los Alamos National Laboratory (LANL) to complete the TEP portion of the project. SRNL's participation builds on the laboratory's decades of work with hydrogen and its isotopes deuterium and tritium - providing the applied research and development that supports the Savannah River Site's handling of tritium. SRNL's experience and expertise in large-scale tritium processing systems and its track record of effective project execution are a unique combination that is key to the success of the ITER project. LANL brings to the partnership experience and expertise in tritium processing technologies specific to the fusion program. This knowledge and understanding were gained through the development and operation of the Tritium Systems Test Assembly at Los Alamos for over 20 years starting in the late 1970's. The US's implementation of the tokamak exhaust processing (TEP) system will provide a technically mature, robust, and cost-effective solution for the separation of hydrogen isotopes from the tokamak exhaust stream. The TEP technology, design challenges, and project status will be presented. (orig.)

  14. Application of MDSplus on EAST Tokamak

    International Nuclear Information System (INIS)

    EAST is a fully superconducting Tokamak in China used for controlled fusion research. MDSplus, a special software package for fusion research, has been used successfully as a central repository for analysed data and PCS (Plasma Control System) data since the debugging experiment in the spring of 2006. In this paper, the reasons for choosing MDSplus as the analysis database and the way to use it are presented in detail, along with the solution to the problem that part of the MDSplus library does not work in the multithread mode. The experiment showed that the data system based on MDSplus operated stably and it could provide a better performance especially for remote users

  15. Application of MDSplus on EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    QU Lianzheng; LUO Jiarong; LI lingling; ZHANG Mingxing; WANG Yong

    2007-01-01

    EAST is a fully superconducting Tokamak in China used for controlled fusion research. MDSplus, a special software package for fusion research, has been used successfully as a central repository for analysed data and PCS (Plasma Control System) data since the debugging experiment in the spring of 2006 . In this paper, the reasons for choosing MDSplus as the analysis database and the way to use it are presented in detail, along with the solution to the problem that part of the MDSplus library does not work in the multithread mode. The experiment showed that the data system based on MDSplus operated stably and it could provide a better performance especially for remote users.

  16. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Box, F.M.A.; Kolk, E. van de [Associatie Euratom-FOM, Nieuwegein (Netherlands). FOM-Instituut voor Plasmafysica; Howard, J. [Plasma Research Laboratory, Research School of Physical Science and Engineering, Australian National University, Canberra 0200 (Australia); Meijer, F.G. [Physics Faculty, University of Amsterdam, Amsterdam (Netherlands)

    1997-03-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. Several spectrometers, equipped with a charge-coupled device array, are being used with spectral ranges in the visible, the vacuum UV and the extreme UV. (orig.)

  17. Bootstrap current in a tokamak

    International Nuclear Information System (INIS)

    The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and βp must be kept below a critical value

  18. Spheromak injection into a tokamak

    OpenAIRE

    Brown, M R; Bellan, P. M.

    1990-01-01

    Recent results from the Caltech spheromak injection experiment [to appear in Phys. Rev. Lett.] are reported. First, current drive by spheromak injection into the ENCORE tokamak as a result of the process of magnetic helicity injection is observed. An initial 30% increase in plasma current is observed followed by a drop by a factor of 3 because of sudden plasma cooling. Second, spheromak injection results in an increase of tokamak central density by a factor of 6. The high-current/high-density...

  19. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    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*cTe/eB(δni/ni)rms which is also derived by a simple theory, the cross-field diffusion time, tp=a2/D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  20. Stellarator - tokamak configurations

    International Nuclear Information System (INIS)

    The stellarator configuration and tokamak configuration with helical fields have been studied both from an equilibrium and stability point of view. The model was restricted to a surface current model with a sharp boundary between plasma and vacuum. A general derivation of equilibrium and stability based on the Energy Principle is given. Physically the unstable modes are identified as external global modes. Detailed numerical results in different parameter regimes are presented and discussed. Critical β-limits for equilibrium and stability are obtained and in particular it is shown that in certain parameter ranges there exist a high-β as well as a low-β-region of stability. 7 refs., 14 figs

  1. Kinetic analysis of MHD ballooning modes in tokamaks

    International Nuclear Information System (INIS)

    A comprehensive analysis of the stability properties of the appropriate kinetically generalized form of MHD ballooning modes together with the usual trapped-particle drift modes is presented. The calculations are fully electromagnetic and include the complete dynamics associated with compressional ion acoustic waves. Trapped-particle effects along with all forms of collisionless dissipation are taken into account without approximations. The influence of collisions is estimated with a model Krook operator. Results from the application of this analysis to realistic tokamak operating conditions indicate that unstable short-wavelength modes with significant growth rates can extend from β = 0 to value above the upper ideal-MHD-critical-beta associated with the so-called second stability regime. Since the strength of the relevant modes appears to vary gradually with β, these results support a soft beta limit picture involving a continuous (rather than abrupt or hard) modification of anomalous transport already present in low-β-tokamaks. However, at higher beta the increasing dominance of the electromagnetic component of the perturbations indicated by these calculations could also imply significantly different transport scaling properties

  2. New High Resolution Thomson Scattering system for the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Brotánková, Jana; Bělský, Petr; Weinzettl, Vladimír; Böhm, Petr

    Vol. 2. Prague : MATFYZPRESS, Prague, 2007 - (Šafránková, J.; Pavlů, J.), s. 218-223 ISBN 978-80-7378-024-1. [Annual Conference of Doctoral Students - WDS 2007 /16./. Prague (CZ), 05.06.2007-08.06.2007] R&D Projects: GA ČR GD202/03/H162 Institutional research plan: CEZ:AV0Z20430508 Keywords : Thomson Scattering * tokamak * diagnostics * laser * electron temperature * electron density * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/wds07.htm#ppm

  3. U-probe for the COMPASS Tokamak

    Czech Academy of Sciences Publication Activity Database

    Kovařík, Karel; Ďuran, Ivan; Stöckel, Jan; Seidl, Jakub; Šesták, David; Brotánková, J.; Spolaore, M.; Martines, E.; Vianello, N.; Hidalgo, C.; Pedrosa, M. A.

    Prague : MATFYZPRESS, 2011 - (Šafránková, J.; Pavlů, J.), s. 227-232 ISBN 978-80-7378-185-9. - (WDS. 2). [WDS 2011 - Annual Conference of Doctoral Students /20./. Prague (CZ), 31.05.2011-03.06.2011] R&D Projects: GA ČR GD202/08/H057; GA MŠk 7G09042; GA MŠk 7G10072 Grant ostatní: EUROATOM(XE) FU07-CT-2007-00060 Institutional research plan: CEZ:AV0Z20430508 Keywords : edge plasma * filaments * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics

  4. Atomic Beam Probe Diagnostic for COMPASS Tokamak

    Czech Academy of Sciences Publication Activity Database

    Háček, Pavel; Weinzettl, Vladimír; Stöckel, Jan; Anda, G.; Veres, G.; Zoletnik, S.; Berta, M.

    Vol. 2. Prague: MATFYZPRESS, 2010 - (Šafránková, J.; Pavlů, J.), s. 7-11. (WDS'10). ISBN 978-80-7378-140-8. [Annual Conference of Doctoral Students - WDS 2010 /19th./. Prague (CZ), 01.06.2010-04.06.2010] R&D Projects: GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma diagnostics * tokamak * COMPASS * beam diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics http://server.ipp.cas.cz/~vwei/work/wds2010_201_f2.pdf

  5. Progress on Joint Experiments on Small Tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Van Oost, G.; Del Bosco, E.; Berta, M.; Brotánková, Jana; Dejarnac, Renaud; Dufková, Edita; Ďuran, Ivan; Hron, Martin; Zajac, Jaromír; Malaquias, A.; Mank, G.; Peleman, P.; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zoletnik, S.; Tál, B.; Ferreira, J.; Fonseca, A.; Hegazy, H.; Kuznetsov, Y.; Ruchko, L.; Vorobyev, G.M.; Ovsyannikov, A.; Sukhov, E.; Singh, A.; Kuteev, B.; Melnikov, A.; Vershkov, V.; Kirneva, N.; Kirnev, G.; Budaev, V.; Sokolov, M.; Talebitaher, A.; Khorshid, P.; Ramos, G.; El Chama Neto, I.; Kraemer-Flecken, A.W.; Soldatov, V.; Marques Fonseca, A.M.; Gutierrez-Tapia, C.R.; Krupnik, L.I.

    Warsaw: EPS, 2007, P-1.070-P-1.070. (Europhysics Conference Abstracts). ISBN 978-83-926290-0-9. [EPS Conference on Plasma Physics/34th./. Warsaw (PL), 02.07.2007-06.07.2007] R&D Projects: GA AV ČR KJB100430504 Grant ostatní: EU(XE) INTAS 100008-8046 Institutional research plan: CEZ:AV0Z20430508 Source of funding: R - rámcový projekt EK Keywords : tokamak * edge plasma * turbulence * improved confinement * plasma diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics http://www.eps2007.ifpilm.waw.pl/pdf2/P1_070.pdf

  6. First Spectroscopic Measurements on the COMPASS Tokamak

    Czech Academy of Sciences Publication Activity Database

    Naydenkova, Diana; Stöckel, Jan; Weinzettl, Vladimír; Šesták, David; Havlíček, Josef

    Vol. 2. Prague : MATFYZPRESS, Prague, 2009 - (Šafránková, J.; Pavlů, J.), s. 158-162 ISBN 978-80-7378-102-6. [Annual conference of doctoral students - WDS 2009 /18./. Prague (CZ), 02.06.2009-05.06.2009] R&D Projects: GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * tokamak * spectroscopic measurements Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/pdf09/WDS09_227_f2_Naydenkova.pdf

  7. Diagnostic Lithium Beam System for COMPASS Tokamak

    Czech Academy of Sciences Publication Activity Database

    Háček, P.; Weinzettl, Vladimír; Stöckel, Jan; Anda, G.; Veres, G.; Zoletnik, S.; Berta, M.

    Prague : MATFYZPRESS, 2011 - (Šafránková, J.; Pavlů, J.), s. 215-220 ISBN 978-80-7378-185-9. - (WDS. 2). [WDS 2011 - Annual Conference of Doctoral Students /20./. Prague (CZ), 31.05.2011-03.06.2011] R&D Projects: GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma diagnostics * tokamak, COMPASS * beam diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics http:// server .ipp.cas.cz/~vwei/work/wds2010_201_f2.pdf

  8. Tokamak physics experiment: Diagnostic windows study

    International Nuclear Information System (INIS)

    We detail the study of diagnostic windows and window thermal stress remediation in the long-pulse, high-power Tokamak Physics Experiment (TPX) operation. The operating environment of the TPX diagnostic windows is reviewed, thermal loads on the windows estimated, and cooling requirements for the windows considered. Applicable window-cooling technology from other fields is reviewed and its application to the TPX windows considered. Methods for TPX window thermal conditioning are recommended, with some discussion of potential implementation problems provided. Recommendations for further research and development work to ensure performance of windows in the TPX system are presented

  9. Electron cyclotron emission imaging in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Munsat, Tobin; Domier, Calvin W.; Kong, Xiangyu; Liang, Tianran; Luhmann, Jr.; Neville C.; Tobias, Benjamin J.; Lee, Woochang; Park, Hyeon K.; Yun, Gunsu; Classen, Ivo. G. J.; Donne, Anthony J. H.

    2010-07-01

    We discuss the recent history and latest developments of the electron cyclotron emission imaging diagnostic technique, wherein electron temperature is measured in magnetically confined plasmas with two-dimensional spatial resolution. The key enabling technologies for this technique are the large-aperture optical systems and the linear detector arrays sensitive to millimeter-wavelength radiation. We present the status and recent progress on existing instruments as well as new systems under development for future experiments. We also discuss data analysis techniques relevant to plasma imaging diagnostics and present recent temperature fluctuation results from the tokamak experiment for technology oriented research (TEXTOR).

  10. Up-down asymmetric tokamak equilibria with parallel flows

    International Nuclear Information System (INIS)

    A class of exact axisymmetric tokamak equilibria with sheared flows parallel to the magnetic field is constructed, generalizing previous work on the subject (Kuiroukidis 2010 Plasma Phys. Control. Fusion 52 015002). The additional free parameters associated with new terms in the solution make it possible to construct up–down asymmetric configurations with a divertor X-point and desirable values of confinement figures of merit as the safety factor on the magnetic axis and plasma betas; in particular, we construct a number of ITER-pertinent equilibria. Their stability with respect to linear MHD perturbations is also examined by applying a sufficient condition.

  11. Threshold condition for nonlinear tearing modes in tokamaks

    International Nuclear Information System (INIS)

    Low-mode-number tearing, mode nonlinear evolution is analyzed emphasizing the need for a threshold condition, to account for observations in tokamaks. The discussion is illustrated by two models recently introduced in the literature. The models can be compared with the available data and/or serve as a basis for planning some experiments in order to either test theory (by means of beta-limit scaling laws, as proposed in this paper) or attempt to control undesirable tearing modes. Introducing a threshold condition in the tearing mode stability analysis is found to reveal some bifurcation points and thus domains of intrinsic stability in the island dynamics operational space

  12. First-wall sputtering effects in Cat-D tokamaks

    International Nuclear Information System (INIS)

    Cat-D tokamaks offer the important advantages of elimination of the need to breed tritium and less frequent first-wall replacement due to reduced neutron damage effects. To capitalize on these features, however, the erosion of the low-Z first-wall coating must be addressed. As shown here, this appears to be feasible if a highly efficient (beta > 0.9) mechanical limiter system can be developed. Still, due to the many uncertainties involved, the simultaneous development of in situ techniques for redeposition coatings appears desirable

  13. Heart rate distribution and predictors of resting heart rate after initiation of beta-blocker treatment in patients with coronary artery disease: REsults of Sympathetic Evaluation And Research of China(RESEARCH) study

    Institute of Scientific and Technical Information of China (English)

    ZHAO Ying-xin; LI Yue-ping; GAO Fei; MA Han-ying; WANG Zhi-jian; HAN Hong-ya; SHEN Hua

    2013-01-01

    Background The importance of heart rate as secondary prevention strategies for patients with coronary artery disease (CAD) is emphasized by multiple guidelines.However,limited information is available on the heart rate distribution and the change patterns of resting heart rate when initiating beta-blocker therapy among Chinese patients with CAD.Methods The REsults of Sympathetic Evaluation And Research of China (RESEARCH) study is a multi-centre,prospective,observational study involving 147 centers in 23 cities across China.All eligible beta-blocker naive patients were prescribed with metroprolol succinate.Initial dosage and target heart rate were selected at the discretion of their physicians in charge according to their usual institutional practice.The heart rate distribution and the change patterns of resting heart rate after initiation of beta-blocker therapy were observed.Results The majority of patients (63.6%) were prescribed with 47.5 mg metroprolol succinate.At baseline,there were only 17.4% of patients whose heart rate was less than 70 beats per minute,and the proportion reached 42.5% and 79.1%,one month and two months after initiation of beta-blockers,respectively.Multivariate linear regression analysis showed that baseline heart rate (B=0.900,SE=0.006,t=141.787,P<0.0001) and the dosage (B=-0.007,SE=0.002,t=-3.242,P=0.001) were independent predictors of resting heart rate 2 months after beta-blocker therapy.Conclusions Resting heart rate is not optimally controlled in a broadly representative cohort of Chinese outpatients with CAD even after initiation of β-blocker therapy,and baseline heart rate and the dosage of beta-blocker are both independent predictors of resting heart rate after β-blocker therapy.

  14. Study of the heating of tokamaks by high energy ion beams

    International Nuclear Information System (INIS)

    This research program has encompassed a number of design studies for a steady state (or long pulse) Auto-Resonant Accelerator (ARA) capable of producing intense beams of high energy (4-20 MEV) ions suitable for the heating of large tokamak devices. The different research topics addressed have ranged over a number of questions related to the design of the individual elements of the accelerator itself, along with studies of the injection and stripping of the accelerated ions in the tokamak and their subsequent energy deposition in the tokamak plasma

  15. Measurements with an emissive probe in the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Schrittwieser, R.; Adámek, Jiří; Balan, P.; Hron, Martin; Ionita, C.; Jakubka, Karel; Kryška, Ladislav; Martines, E.; Stöckel, Jan; Tichý, M.; Van Oost, G.

    2002-01-01

    Roč. 44, č. 5 (2002), s. 567-578. ISSN 0741-3335 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : CASTOR tokamak, plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.121, year: 2002

  16. JOINT EXPERIMENTS ON SMALL TOKAMAKS: EDGE PLASMA STUDIES ON CASTOR

    Czech Academy of Sciences Publication Activity Database

    Van Oost, G.; Berta, M.; Brotánková, Jana; Dejarnac, Renaud; Del Bosco, E.; Dufková, Edita; Ďuran, Ivan; Gryaznevich, M.P.; Horáček, Jan; Hron, Martin; Malaquias, A.; Mank, G.; Peleman, P.; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zoletnik, S.; Tál, B.; Ferrera, J.; Fonseca, A.; Hegazy, H.; Kuznetsov, Y.; Ossyannikov, A.; Singh, A.; Sokholov, M.; Talebitaher, A.

    2007-01-01

    Roč. 47, č. 5 (2007), s. 378-386. ISSN 0029-5515 R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * edge plasma * turbulence * Langmuir probe * plasma radiation * Hall probe Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.278, year: 2007

  17. Technology and physics in the Tokamak Program: The need for an integrated, steady-state RandD tokamak experiment

    International Nuclear Information System (INIS)

    The Steady-state Tokamak (STE) Experiment is a proposed superconducting-coil, hydrogen-plasma tokamak device intended to address the integrated non-nuclear issues of steady state, high-power tokamak physics and technology. Such a facility has been called for in the US program plan for the mid 1990's, and will play a unique role in the world-wide fusion effort. Information from STE on steady-state current drive, plasma control, and high power technology will contribute significantly to the operating capabilities of future steady-state devices. This paper reviews preliminary designs and expected technological contributions to the US and world fusion reactor research from each of the above mentioned reactor systems. This document is intended as a proposal and feasibility discussion and does not include exhaustive technical reviews. 12 figs., 3 tabs

  18. Development of large insulator rings for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    This paper discusses research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applications, fabrication approach and testing activities are highlighted

  19. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  20. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    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

  1. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  2. The design of the Tokamak Physics Experiment (TPX)

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, J.A.; Goldston, R.J.; Sinnis, J.C.; Bernabei, S.; Bialek, J.M.; Bronner, G.; Chen, S.J.; Chrzanowski, J.; Citrolo, J.; Dahlgren, F.

    1993-09-01

    The Tokamak Physics Experiment (TPX) is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a low-activation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation of US industry. Operation will begin with first plasma in the year 2000.

  3. MHD stability limits in the TCV Tokamak

    International Nuclear Information System (INIS)

    of this limit with elongation is also in qualitative agreement with ideal MHD theory. Edge localised modes (ELMs), occurring in TCV Ohmic high-confinement mode discharges, were observed to be preceded by coherent magnetic oscillations. The detected poloidal and toroidal mode structures are consistent with a resonant flux surface close to the plasma edge. Unlike conventional MHD modes, these precursors start at a random toroidal location and then grow in amplitude and toroidal extent until they encompass the whole toroidal circumference. Thus, the asymmetry causing and maintaining the toroidal localisation of the ELM precursor must be intrinsic to the plasma. Soft X-ray measurements show that the localised precursor always coincides with a central m = 1 mode, which can usually be associated with the sawtooth pre- or postcursor mode. A comparison of the phases indicates a correlation with the maximum of the central mode preceding the toroidal location of the ELM precursor and, therefore, a hitherto unobserved coupling between central modes and ELMs. Highly elongated plasmas promise several advantages, among them higher current and beta limits. During TCV experiments dedicated to an increasing of the plasma elongation, a new disruptive current limit, at values well below the conventional current limit corresponding to qa > 2, was encountered for κ > 2.3. This limit, which is preceded by a kink-type mode, is found to be consistent with ideal MHD stability calculations. The TCV observations, therefore, provide the first experimental confirmation of a deviation of the linear Troyon-scaling of the ideal beta limit with normalised current at high elongation, which was predicted over 10 years ago. Neoclassical tearing modes (NTMs), which have been observed to limit the achievable beta in a number of tokamaks, arise from a helical perturbation of the bootstrap current caused by an existing seed island. Neoclassical m/n = 2/1 tearing modes have been identified in TCV

  4. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

    observed decrease of this limit with elongation is also in qualitative agreement with ideal MHD theory. Edge localised modes (ELMs), occurring in TCV Ohmic high-confinement mode discharges, were observed to be preceded by coherent magnetic oscillations. The detected poloidal and toroidal mode structures are consistent with a resonant flux surface close to the plasma edge. Unlike conventional MHD modes, these precursors start at a random toroidal location and then grow in amplitude and toroidal extent until they encompass the whole toroidal circumference. Thus, the asymmetry causing and maintaining the toroidal localisation of the ELM precursor must be intrinsic to the plasma. Soft X-ray measurements show that the localised precursor always coincides with a central m = 1 mode, which can usually be associated with the sawtooth pre- or postcursor mode. A comparison of the phases indicates a correlation with the maximum of the central mode preceding the toroidal location of the ELM precursor and, therefore, a hitherto unobserved coupling between central modes and ELMs. Highly elongated plasmas promise several advantages, among them higher current and beta limits. During TCV experiments dedicated to an increasing of the plasma elongation, a new disruptive current limit, at values well below the conventional current limit corresponding to q{sub a} > 2, was encountered for {kappa} > 2.3. This limit, which is preceded by a kink-type mode, is found to be consistent with ideal MHD stability calculations. The TCV observations, therefore, provide the first experimental confirmation of a deviation of the linear Troyon-scaling of the ideal beta limit with normalised current at high elongation, which was predicted over 10 years ago. Neoclassical tearing modes (NTMs), which have been observed to limit the achievable beta in a number of tokamaks, arise from a helical perturbation of the bootstrap current caused by an existing seed island. Neoclassical m/n = 2/1 tearing modes have been

  5. Joint Experiments on the Tokamaks CASTOR and T-10

    International Nuclear Information System (INIS)

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confiement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. The first Joint (Host Laboratory) Experiment (JE1) has been carried out in 2005 on the CASTOR tokamak at the IPP Prague, Czech Republic. It was jointly organized by the IPP-ASCR and KFKI HAC, Budapest, involved 20 scientists from 7 countries and was supported through the IAEA and the ICTP, Trieste. The objective of JE1 was to perform studies of plasma edge turbulence and plasma confinement. Following the success of JE1, JE2 has been performed on T-10 at RRC 'Kurchatov Institute' in Moscow; 30 scientists from 13 countries participated in this experiment. This experiment aimed to continue JE1 turbulence studies, now extending them to the plasma core. Results of JE1 and JE2 will be overviewed and compared

  6. Interleukin 1-beta and the research process of periodontal disease%白细胞介素1-β与牙周炎的研究进程

    Institute of Scientific and Technical Information of China (English)

    曾毅

    2013-01-01

    1-βinterleukin (IL-1 beta) is a kind of important immune cellfactors in the main genetic interleukin 1 of the members of the family. In recent years, study found that IL-1 beta can be immune to gum disease occurrence, development plays an important role, and found that IL-1βgene polymorphism and on the type of periodontal disease, the in-depth study of the IL-1 beta, and the relationship between periodontal disease, the cause of periodontal disease research, prevention, treatment and prognosis judgment have far-reaching significance.%白细胞介素1-β(IL-1β)是一类重要的免疫细胞因子白细胞介素1的主要基因家族成员之一,近年来研究发现IL-1β在牙周病的免疫发生、发展中发挥着重要的作用,并且发现IL-1β基因多态性与牙周炎的类型有关,深入研究IL-1β和牙周炎的关系,对牙周炎的病因研究、预防、治疗以及预后判断有深远意义。

  7. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

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

  8. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (ne and Te) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  9. Controlling fusion yield in tokamaks with spin polarized fuel, and feasibility studies on the DIII-D tokamak

    International Nuclear Information System (INIS)

    The march towards electricity production through tokamaks requires the construction of new facilities and the inevitable replacement of the previous generation. There are, however, research topics that are better suited to the existing tokamaks, areas of great potential that are not sufficiently mature for implementation in high power machines, and these provide strong support for a balanced policy that includes the redirection of existing programs. Spin polarized fusion, in which the nuclei of tokamak fuel particles are spin-aligned and favorably change both the fusion cross-section and the distribution of initial velocity vectors of charged fusion products, is described here as an example of a technological and physics topic that is ripe for development in a machine such as the DIII-D tokamak. In this study, such research and development experiments may not be efficient at the ITER-scale, while the plasma performance, diagnostic access, and collaborative personnel available within the United States' magnetic fusion research program, and at the DIII-D facility in particular, provide a unique opportunity to further fusion progress

  10. Industry roles in the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    There are several distinguishing features of the Tokamak Physics Experiment (TPX) to be found in the TPX program and in the organizations for constructing and operating the machine. Programmatically, TPX addresses several issues critical to the viability of magnetic fusion power plants. Organizationally, it is a multi-institutional partnership to construct and operate the machine and carry out its program mission. An important part of the construction partnership is the integrated industrial responsibility for design, R ampersand D, and construction. The TPX physics design takes advantage of recent research on advanced tokamak operating modes achieved for time scales of the order of seconds that are consistent with continuous operation. This synergism of high performance (higher power density) modes with plasma current driven mostly by internal pressure (boot-strap effect) points toward tokamak power plants that will be cost-competitive and operate continuously. A large fraction of the project is subcontracted to industry. By policy, these contracts are at a high level in the project breakdown of work, giving contractors much of the overall responsibility for a given major system. That responsibility often includes design and R ampersand D in addition to the fabrication of the system in question. Each contract is managed through one of three national laboratories: PPPL, LLNL, and ORNL. Separate contracts for system integration and construction management round out the industry involvement in the project. This integrated, major responsibility attracts high-level corporate attention within each company, which are major corporations with long-standing interest in fusion. Through the contracts already established on the TPX project, a new standard for industry involvement in fusion has been set, and these industries will be well prepared for future fusion projects

  11. High performance advanced tokamak regimes in DIII-D for next-step experiments

    International Nuclear Information System (INIS)

    Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] 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 stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric 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 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

  12. Overview on Chinese tokamak experimental progress

    International Nuclear Information System (INIS)

    Tokamak experiment research in China has made important progress. The main efforts subjected to quasi-steady state operation, LHCD, plasma heating with ICRF, IBW, NBI, ECRH, fueling with pellet and supersonic molecular beam, first wall conditioning technique. Plasma parameters in experiments were much improved, such as ne=8x1019m-3, plasma pulse >10Sec. ICRF boronization and conditioning made Zeff close to unit. Steady state full LH wave current drive has been achieved for more than 3 seconds. LHCD ramp up and recharge have also been demonstrated. The Best ηCDexp∼0.5(1+0.085 exp(4.8(BT-1.45))neICDRp/PLH=1019m-2A/W. Quasi steady state H-mode like plasma with density close to Greenwald limit was obtained by LHCD, in which energy confinement time was nearly 5 times longer than the Ohmic case. The synergy between IBW, pellet and LHCD was tested. Research on the mechanism of macro-turbulence has been extensively carried out experimentally. Ac operation of tokamak was successfully demonstrated. (author)

  13. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  14. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  15. Theory and calculation of finite beta drift wave turbulence

    International Nuclear Information System (INIS)

    Using numerical techniques, we calculate eigenmodes of the nonlinear universal mode with finite beta in order to determine the scaling of the saturation level of the instability with beta. We use two different renormalizations in the calculations and find that using the appropriate renormalization, we are able to recover Alcator density scaling, as originally found in analytic work by Molvig and Hirshman. We also find that the universal mode should be stable in ohmically heated tokamaks above a critical beta on the order of 0.02

  16. Decrease of Hepatitis C Burden in Patients With Transfusion Dependent Beta Thalassemia Major, Thalassemia Research Center, 1995 – 2014

    Directory of Open Access Journals (Sweden)

    Mehrnoush Kosaryan

    2015-07-01

    Full Text Available Background: Chronic hepatitis C infection used to be one of the most important burdens on patients with transfusion-dependent beta thalassemia major (TDTM. Chronic active hepatitis reduces quality of life, and liver cirrhosis and cancer shorten life expectancy in many cases. Objectives: We compared the characteristics of our patients at the Thalassemia Research Center (TRC regarding hepatitis C infection at two time points. Patients and Methods: A review was conducted in a cohort of 390 TDTM patients with a history of at least one blood transfusion in 2014. Type of treatment protocol for hepatitis C virus (HCV and the number of courses were defined. Descriptive statistics were performed using SPSS software (V16. Results: Screening for HCV started in 1995 at the TRC. Seventy-seven (15% patients were antibody-positive in 1995. Tests for virus detection were not available at the time. Patients have been examined using serum AST, ALT, bilirubin, PT, PTT, and liver biopsy, and 45 were treated using alpha interferon alone. A second liver biopsy was performed at the end of treatment for 21 patients, and a blinded pathologist compared the histology according to the Knodell score. According to normalization of liver enzymes, the treatment was successful (McNemar test, P < 0.02. Based on the Knodell score, 54%, 31%, and 11% had complete, partial, and no response, respectively. A quantitative test for viremia became available thereafter. Thirteen patients who were resistant to alpha interferon have been treated using “Pegasys”™ ± ribavirin. Ten patients responded; however, three have been resistant and are still viremic. Twenty-seven patients received no treatment. Twenty-two (81.4% had negative PCR tests. Five viremic patients refused treatment. A second screening test for HCV antibody was introduced in 2001, and, since then, annual screening for HCV antibody has been performed for all patients. No new case has been found since 2001. During the

  17. Development path of low aspect ratio tokamak power plants

    International Nuclear Information System (INIS)

    Recent advances in tokamak physics indicate that a spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants. Full calculations of stability to kink and ballooning modes show the possibility of greater than 50% β-toroidal with the normalized β [βN=βT/(I/ab)] as high as 10 and fully aligned 100% bootstrap current. Such β-values coupled with 2-3 T toroidal fields imply a pilot plant about the size of the present DIII-D tokamak could produce ∝800 MW thermal, 160 MW net electric, and would have a ratio of gross electric power over recirculating power (QPLANT) of 1.9. The high β values in the ST mean that E x B shear stabilization of turbulence should be ten times more effective in the ST than in present tokamaks, implying that the required high quality of confinement needed to support such high beta values will be obtained. The anticipated β values are so high that the allowable neutron flux at the blanket sets the device size, not the physics constraints. The ST has a favorable size scaling so that at 2-3 times the pilot plant size the QPLANT rises to 4-5, an economic range and 4 GW thermal power plants result. Current drive power requirements for 10% of the plasma current are consistent with the plant efficiencies quoted. The unshielded copper centerpost should have an adequate lifetime against nuclear transmutation induced resistance change and the low voltage, high current power supplies needed for the 12 turn TF coil appear reasonable. (orig.)

  18. Conceptual design of a commercial tokamak reactor using resistive magnets

    International Nuclear Information System (INIS)

    The potential of resistive magnet tokamaks as commercial electricity producing power plants is investigated. Parametric studies indicate that attractive design space exists for these reactors at relatively low field (2.5 to 4.5 T), moderate wall loading (3 to 4 MW/m2) and medium to large net electric outputs (>600 MW/sub e/). High toroidal beta (20 to 25%) possible in the second regime of plasma stability may provide advantages of reduced recirculating power and plasma current but moderate beta reactors (6 to 10%) remain attractive. A conceptual design for the Resistive magnet Commercial Tokamak Reactor (RCTR) is presented. The layout of the nuclear island is driven by compatibility requirements of the demounting capability with structural and blanket design considerations. The nuclear island is fully demountable with access to all components within the toroidal field coils possible via simple vertical lifts. The blanket system, segmented for vertical removal, uses a self-cooled liquid lithium breeder/coolant with vanadium structure and an HT-9 reflector. The first wall is also lithium cooled with a vanadium structure but is constructed in a single, pre-tested unit for assembly and periodic replacement. Ohmic and equilibrium field-coils are located within the bore of the toroidal field coil for improved performance

  19. Second regime tokamak operation at large aspect ratio

    International Nuclear Information System (INIS)

    The equilibrium, stability, and transport properties of large aspect ratio tokamaks operating at the second stability regime are described theoretically using numerical and analytical techniques. It has been shown that, at large aspect ratio, significant current profile control is possible with relatively modest amounts of neutral beam current drive, and the power needed to access and maintain the second regime operation is calculated to be about 3 MW using the results of an integrated 1.5D transport and stability code. An example second regime experiment has been described and the results are presented of extensive calculations illustrating several possible operating scenarios, external and internal model stability boundaries, and the experimental features needed to evaluate and test the high beta tokamak theories. The theory which describes the stabilizing effect of energetic particles during high beta operation was extended to finite aspect ratio. A key technical problem for application of this technique appears to be caused by ripple transport. Plasma rotation effects are found to be generally destabilizing and several other schemes for improved access to the second stability regime are discussed including ponderomotive stabilization of the plasma edge region and active feedback control

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

    International Nuclear Information System (INIS)

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

  1. RF preionization in Tokamak thor

    International Nuclear Information System (INIS)

    During the study of the RF preionization in Tokamak Thor was observed that the starting of the plasma and its time behaviour were correlated with the presence of resonance conditions both at the electron cyclotron frequency Ωsub(deg) and at its sub-harmonics Ωsub(deg)/n. These results are supported by a simple qualitative calculation

  2. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

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

  3. 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

  4. ICRF heating experiments in JFT-2 tokamak

    International Nuclear Information System (INIS)

    This is an experimental study of ICRF heating on JFT-2 Tokamak in Japan Atomic Energy Research Institute. In this study, we first clarified physical and engineering problems of ICRF heating of tokamak plasma. Next, we optimized the design of the ICRF heating system, and the plasma parameters for the heating. Finally, we could demonstrate a high efficiency of this additional heating method by launching RF power which is two or three times as large as an ohmic input power to a plasma. And we achieved following things. (1) We optimized a design of an antenna, and we improved a durability of the system for high voltage. With the result that we achieved the maximum power density on an antenna. (2) We demonstrated that electron heating regime and ion heating regime can be easily accessed by controlling plasma parameters. Also we found the optimum heating conditions in each heating regime. (3) We experimentally clarified the production mechanism of impurities during ICRF heating. We could reduce the influx of metal impurity ions to a plasma by employing low z materials for limiters and antenna shields. Consequently, we improved a heating efficiency of electrons. Next, we studied a power balance of plasma during ICRF heating, and we could compare heating characteristics of ICRF with other additional heatings on JFT-2. (author)

  5. Ion cyclotron system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H

    1998-05-01

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs.

  6. Ion cyclotron system design for KSTAR tokamak

    International Nuclear Information System (INIS)

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs

  7. Integrated plasma control for high performance tokamaks

    International Nuclear Information System (INIS)

    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)

  8. Plasma transport in a Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Nominal predicted plasma conditions in a Compact Ignition Tokamak (CIT) are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models that have given almost equally good fits to experimental data. Using a transport model that best fits the data, thermonuclear ignition occurs in a CIT design with a major radius of 1.32 m, plasma half-width of 0.43 mn, elongation of 2.0, and toroidal field and plasma current ramped in 6 s from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the /sup 3/He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates have a large effect on ignition and on the maximum beta that can be achieved

  9. Physics aspects of the compact ignition tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve nτE ≅ 2 x 1020 s/m3 required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which provides a high level of ohmic heating, improves the energy confinement, and allows a relatively high beta (≅ 6%). The present CIT design also has a high degree of elongation (κ ≅ 1.8) to aid in producing the large plasma current. A double null poloidal divertor and pellet injection are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Auxiliary heating is expected to be necessary to achieve ignition, and 10-20 MW of ICRF is to be provided. (orig.)

  10. Assembly of Aditya upgrade tokamak

    International Nuclear Information System (INIS)

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a tokamak with divertor configuration. At present the existing ADITYA tokamak has been dismantled up to bottom plinth on which the whole assembly of toroidal field (TF) coils and vacuum vessel rested. The major components of ADITYA machine includes 20 TF coils and its structural components, 9 Ohmic coils and its clamps, 4 BV coils and its clamps as well as their busbar connections, vacuum vessel and its supports and buckling cylinder, which are all being dismantled. The re-assembly of the ADITYA Upgrade tokamak started with installation and positioning of new buckling cylinder and central solenoid (TR1) coil. After that the inner sections of TF coils are placed following which in-situ winding, installation, positioning and support mounting of two pairs of new inner divertor coils have been carried out. After securing the TF coils with top I-beams the new torus shaped vacuum vessel with circular cross-section in 2 halves have been installed. The assembly of TF structural components such as top and bottom guiding wedges, driving wedges, top and bottom compression ring, inner and outer fish plates and top inverted triangle has been carried out in an appropriate sequence. The assembly of outer sections of TF coils along with the proper placements of top auxiliary TR and vertical field coils with proper alignment and positioning with the optical metrology instrument mainly completes the reassembly. Detailed re-assembly steps and challenges faced during re-assembly will be discussed in this paper. (author)

  11. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Low cost tokamak fusion power reactor based on a reversed shear plasma: CREST

    International Nuclear Information System (INIS)

    The Cost down of tokamak fusion reactors is one of the key issues that must be dealt with in order to deploy fusion energy for commercial use. Our previous study of the costs of tokamak power plants has shown that a very high Troyon coefficient βN (up to 5 or 6) is required in order to realize a cost-competitive tokamak reactor. A reversed shear configuration, which has been observed recently in several tokamak experiments, might results in such a low cost tokamak. In this study, we propose a compact commercial reactor based on the shear reversed high beta equilibrium, which is named the Compact REversed Shear Tokamak (CREST). The optimized parameters of CREST are; major radius R = 5.4 m, maximum toroidal field Bmax = 12.5 T (5.6 T at plasma center), and fusion power Pf = 3 GW with 4.5 MW/m2 in the mean neutron wall load. The ideal MHD instabilities are all stable in this equilibrium while the βN value reaches 5.5. The plasma configuration of CREST is close to that of ITER advanced mode plasma. This will encourage an investigation of the feasibility of the CREST concept by ITER. Although many further studies and developments are necessary, such compact tokamak can be cost-competitive as an electric power source in the 21st century and could be a promising tool in the development of a commercial tokamak reactor succeeding the ITER project. (author)

  12. Conceptual design of Remote Control System for EAST tokamak

    International Nuclear Information System (INIS)

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication

  13. Conceptual design of Remote Control System for EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

    2014-05-15

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

  14. Neutronics design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

  15. Stabilization of external modes in tokamaks by resistive walls and plasma rotation

    International Nuclear Information System (INIS)

    It is shown that low-n pressure driven external modes in tokamaks can be fully stabilized by resistive walls in combination with sonic rotation of the plasma. The stabilization depends on the excitation of sound waves by the toroidal coupling to Alfven waves and is affected by ion Landau damping. Two-dimensional stability calculations are presented to show the gains in the beta limit resulting from this wall stabilization. (author) 4 figs., 13 refs

  16. Transport and effects of high-energy charged fusion-products in tokamaks

    International Nuclear Information System (INIS)

    An explicit expression that describes the guiding-center motion of a high-energy charged particle in a lossless, axisymmetric, low-beta (β less than or equal to 0.1) tokamak has been derived. The expression is valid in any type of toroidal and poloidal field configurations as long as the poloidal flux surfaces, specified by psi (r), are circular. A similar explicit expression is obtained for fields having a vertical stabilizing component, B/sub v/

  17. The cryostat of the CUORE Project, a 1-ton scale cryogenic experiment for Neutrinoless Double Beta Decay Research

    International Nuclear Information System (INIS)

    CUORE is a new generation of 1-ton scale cryogenic detector for rare-events physics. CUORE, a detector to search Neutrinoless Double Beta Decay of 130Te, is an array of 988 TeO2 crystals of a mass of 750 g each. To build the cryogenic system, where the CUORE detector will be installed in the Gran Sasso Underground Laboratory, is really a challenge. It is a large cryogen-free cryostat cooled by pulse tubes and by a high power dilution refrigerator. To avoid radioactive background, about 10000 kg of lead will be cooled to below 1 K and only few construction materials are acceptable. the detector will have a total mass of about 1500 kg and must be cooled to less than 10 mK in a vibration-free environment.

  18. The cryostat of the CUORE Project, a 1-ton scale cryogenic experiment for Neutrinoless Double Beta Decay Research

    Energy Technology Data Exchange (ETDEWEB)

    Schaeffer, David; Nucciotti, Angelo [Dip. di Fisica ' G. Occhialini' , Univ and INFN Sez di Milano-Bicocca, Milano (Italy); Alessandria, Franco [INFN Sez di Milano, Milano (Italy); Ardito, Raffaele [Dip. di Ingegneria Strutturale, Politecnico di Milano, Milano (Italy); Barucci, Marco; Risegari, Lara; Ventura, Guglielmo [Dip. di Fisica, Univ di Firenze and INFN Sez di Firenze, Firenze (Italy); Bucci, Carlo [INFN Laboratori Nazionali del Gran Sasso, Assergi (Italy); Frossati, Giorgio [Leiden Cryogenics, Leiden (Netherlands); Olcese, Marco [INFN Sez di Genova, Genova (Italy); Waard, Arlette de [Kamerlingh Onnes Laboratory, Leiden University, Leiden (Netherlands)], E-mail: david.schaeffer@mib.infn.it

    2009-02-01

    CUORE is a new generation of 1-ton scale cryogenic detector for rare-events physics. CUORE, a detector to search Neutrinoless Double Beta Decay of {sup 130}Te, is an array of 988 TeO{sub 2} crystals of a mass of 750 g each. To build the cryogenic system, where the CUORE detector will be installed in the Gran Sasso Underground Laboratory, is really a challenge. It is a large cryogen-free cryostat cooled by pulse tubes and by a high power dilution refrigerator. To avoid radioactive background, about 10000 kg of lead will be cooled to below 1 K and only few construction materials are acceptable. the detector will have a total mass of about 1500 kg and must be cooled to less than 10 mK in a vibration-free environment.

  19. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    Science.gov (United States)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  20. Annual report of the Fusion Research Center for the period of April 1, 1983 to March 31, 1984

    International Nuclear Information System (INIS)

    Research and development activities of the Fusion Research Center (Department of Thermonuclear Fusion Research and Department of Large Tokamak Development) from April 1983 to March 1984 are described. Installation and commissioning of the new tokamak JFT-2M had been completed. The 2nd ICRF heating experiment and LH current drive experiment were started. In the field of plasma theory, the scaling law of the critical beta in a tokamak was obtained and the ICRF heating was analyzed in detail. The first phase of the cooperation of Doublet III will be finished in Sept. 1984. The JT-60 program progressed as scheduled. Installation of the tokamak machine, initiated in Feb. 1983, will be finished in Sept. 1984. The tests of power supply and control system on site and the fabrication of the neutral beam injectors in factory proceeded successfully. Performance tests of prototype injector unit for JT-60 NBI progressed as scheduled. A new advanced source plasma generator was developed to provide a high proton ratio exceeding 90%. Klystrons for JT-60 LH heating achieved the output power of 1 MW for 10 sec. Performance tests of titanium evaporators for JT-60 were completed. The Japanese coil for IEA Large Coil Task was installed in a test facility at ORNL and the partial cool-down was carried out. Construction of the Tritium Process Laboratory was completed. Design studies of the Fusion Experimental Reactor (FER) and INTOR proceeded. (author)

  1. Transport of Dust Particles in Tokamak Devices

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A Y; Smirnov, R D; Krasheninnikov, S I; Rognlien, T D; Rozenberg, M

    2006-06-06

    Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.

  2. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

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

  3. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

    2012-10-15

    steady-state operating lithium divertor module project for Kazakhstan tokamak KTM. At present the lithium divertor module for KTM tokamak is under development in the framework of ISTC project no. K-1561. Initial heating up to 200 Degree-Sign C and lithium surface temperature stabilization during plasma interaction in the range of 350-550 Degree-Sign C will be provided by external system for thermal stabilization due to circulation of the Na-K heat transfer media. Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Development, creation and experimental research of lithium divertor model for KTM will allow to solve existing problems and to fulfill the basic approaches to designing of lithium divertor and in-vessel elements of new fusion reactor generation, to investigate plasma physics aspects of lithium influence, to develop technology of work with lithium in tokamak conditions. Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation.

  4. Bootstrap Current in Spherical Tokamaks

    Institute of Scientific and Technical Information of China (English)

    王中天; 王龙

    2003-01-01

    Variational principle for the neoclassical theory has been developed by including amomentum restoring term in the electron-electron collisional operator, which gives an additionalfree parameter maximizing the heat production rate. All transport coefficients are obtained in-cluding the bootstrap current. The essential feature of the study is that the aspect ratio affects thefunction of the electron-electron collision operator through a geometrical factor. When the aspectratio approaches to unity, the fraction of circulating particles goes to zero and the contribution toparticle flux from the electron-electron collision vanishes. The resulting diffusion coefficient is inrough agreement with Hazeltine. When the aspect ratio approaches to infinity, the results are inagreement with Rosenbluth. The formalism gives the two extreme cases a connection. The theoryis particularly important for the calculation of bootstrap current in spherical tokamaks and thepresent tokamaks, in which the square root of the inverse aspect ratio, in general, is not small.

  5. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell

  6. Frascati Tokamak transformer switching system

    International Nuclear Information System (INIS)

    Plasma ionization and heating, in the Frascati Tokamak, is obtained generating an emf along the plasma column, by switching the dc current flowing in the Tokamak transformer. 30 kA flowing in the 60 mH transformer inductance must be commutated into a resistance to generate 40 kV across the transformer itself. Studies and tests to solve this problem have been conducted, on different types of breakers, in cooperation between Tecnomasio Italiano Brown Boveri, Milan and Laboratori Gas Ionizzati, Frascati. Satisfactory results have finally been obtained using a DLF commercial air blast breaker in a chopper type circuit. A capacitor bank in parallel to the breaker is discharged immediately after the contacts separation and the arc in the switching element is extinguished at the first current zero. A saturable reactance in series with the breaker reduces the current decay rate to allow sufficient deionization time

  7. Burn Control Mechanisms in Tokamaks

    Science.gov (United States)

    Hill, Maxwell; Stacey, Weston

    2013-10-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamaks, especially those used as a neutron source for fusion-fission hybrid reactors, such as the Subcritical Advanced Burner Reactor (SABR) concept. At Georgia Tech, we are developing a new burning plasma dynamics code to investigate passive safety mechanisms that could prevent power excursions in tokamak reactors. This code solves the coupled set of balance equations governing burning plasmas in conjunction with a two-point SOL-divertor model. Predictions have been benchmarked against data from DIII-D. We are examining several potential negative feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instabilities, iii) the degradation of alpha-particle confinement resulting from ripples in the toroidal field, iv) modifications to the radial current profile, v) ``divertor choking'' and vi) Type 1 ELMs.

  8. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Jae; Cheong, Y. H.; Choi, Y. S.; Cheon, E. J. [PlaGen, Seoul (Korea, Republic of)

    2003-11-15

    The establishment of regulatory and approval systems for KSTAR (Korea Superconducting Tokamak Advanced Research) has been demanded as the facility is targeted to be completed in the year of 2005. Such establishment can be achieved by performing adequate and in-depth analyses on safety issues covering radiological and chemical hazard materials, radiation protection, high vacuum, very low temperature, etc. The loss of coolant accidents and the loss of vacuum accident in fusion facilities have been introduced with summary of simulation results that were previously reported for ITER and JET. Computer codes that are actively used for accident simulation research are examined and their main features are briefly described. It can be stated that the safety analysis is indispensable to secure the safety of workers and individual members of the public as well as to establish the regulatory and approval systems for KSTAR tokamak.

  9. Equilibrium Reconstruction in EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    QIAN Jinping; WAN Baonian; L. L. LAO; SHEN Biao; S. A. SABBAGH; SUN Youwen; LIU Dongmei; XIAO Singjia; REN Qilong; GONG Xianzu; LI Jiangang

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of toka-mak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier ex-pansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.

  10. Instrumentation and controls of an ignited tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega R.J.; Raju, G.V.S.; Stone, R.S.

    1980-10-01

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented.

  11. Instrumentation and controls of an ignited tokamak

    International Nuclear Information System (INIS)

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented

  12. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

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

    Energy Technology Data Exchange (ETDEWEB)

    Castracane, J.

    2001-01-04

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.

  14. Multi-mode remote participation on the GOLEM tokamak

    Czech Academy of Sciences Publication Activity Database

    Svoboda, V.; Huang, B.; Mlynář, Jan; Pokol, G.I.; Stöckel, Jan; Vondrášek, G.

    2011-01-01

    Roč. 86, 6-8 (2011), s. 1310-1314. ISSN 0920-3796. [Symposium on Fusion Technology (SOFT) /26th./. Porto, 27.09.2010-01.10.2010] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * remote participation * education Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.490, year: 2011 http://www.sciencedirect.com/science/article/pii/S0920379611002390

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

    International Nuclear Information System (INIS)

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies

  16. Optimization of magnetic perturbation spectra for the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Cahyna, Pavel

    Geneva : International Atomic Energy Agency, 2008. s. 247-247. ISBN N. [IAEA Fusion Energy Conference/22nd./. 13.10.2008-18.10.2008, Geneva] Institutional research plan: CEZ:AV0Z20430508 Keywords : resonant magnetic perturbations * ELM control * magnetic islands * saddle coils * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www-pub.iaea.org/MTCD/Meetings/PDFplus/2008/cn165/cn165_BookOfAbstracts.pdf

  17. Plasma edge biasing on CASTOR tokamak using LHCD

    Czech Academy of Sciences Publication Activity Database

    Žáček, František; Petržílka, Václav; Jakubka, Karel; Stöckel, Jan; Gunn, J.; Goniche, M.; Devynck, P.; Podesta, M.; Nanobashvili, S.

    2001-01-01

    Roč. 51, č. 10 (2001), s. 1129-1138. ISSN 0011-4626. [Europhysics Workshop on Role of Electric Fields in Plasma Confinement and Exhaust/4th./. Funchal, Madeira, 24.06.2001-25.06.2001] R&D Projects: GA AV ČR IAA1043101 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.345, year: 2001

  18. A predictive study of non-inductive current driven KSTAR tokamak discharge modes using a new transport/heating simulation package

    International Nuclear Information System (INIS)

    For the physical objectives of the KSTAR tokamak, a predictive transport/heating/current drive simulation package has been developed and named ASTRA/LSC/TORIC. A new simulation package consists of the ray tracing lower hybrid wave code LSC, a neutral beam Fokker-Planck solver and the ICRF full wave code TORIC together with the transport code ASTRA. The new simulation package ASTRA/LSC/TORIC can be applied to many interesting research areas such as heating and non-inductive current drive physics for a steady state plasma, predictive study of advanced tokamak modes with a, for example, high beta, high H factor and high bootstrap current. Studies into two kinds of question are made: (a) Can the total required plasma current Ip (a baseline heating mode with 1.5 MA and an upgrade heating mode with 2.0 MA) be driven fully non-inductively and be sustained during a flat-top phase of reasonable KSTAR reference plasma parameter ranges with the help of auxiliary KSTAR heating (H) and current drive (CD) systems (NBH/CD, LHH/CD and ICRH/CD)? (b) Can a reversed shear mode be developed in an early ramp-up phase and be sustained during a flat-top phase? These predictive simulation results using a newly developed simulation package are presented. (author)

  19. Metrology measurements for Aditya tokamak upgradation

    International Nuclear Information System (INIS)

    After 25 years of Aditya tokamak (midsized, air-core, R0= 75 cm, a = 25 cm) operation achieving high temperature circular plasmas in limiter configuration, upgrading it to Aditya-U tokamak with divertor configuration has been planned and the upgradation is under progress. The upgradation process include dismantling of the existing Aditya tokamak to its base level and re-erect it by placing new subsystems like new vacuum vessel of circular cross-section, new buckling cylinder etc. Apposite alignment of subsystems, mainly all the magnetic coil systems in all grades and scales of tokamak is very crucial and essential, as misaligned magnetic coil system scan generate error magnetic fields, which can significantly impact the plasma formation and sustainment in a tokamak. With this motivation, position and alignment measurement of the existing magnetic coils and structural components of ADITYA tokamak is carried out for the very first time with the optical metrology instrument. Prior to carrying out measurement exercise, machine datum has been transferred to the reference on the wall of tokamak hall using five-point laser and the machine center has been transformed to the four wall of tokamak hall. All position measurements are done with respect to machine major axis in cylindrical geometry. Measurement includes existing radial (R) and elevation (Z) positions of all magnetic coils and various structural components within the accuracy of ± 1 mm. More than 5000 data points are recorded using optical metrology instrument. Again the elevation references are transferred to the primary network established and the angular references are transformed on the floor of the tokamak hall. These results will serve as ready reference for reassembly and alignment of Aditya - Upgrade tokamak. In this paper detailed position measurements of different subsystems of old Aditya tokamak and the relocation of them along with new ones using the optical metrology instruments will be presented

  20. KTM Tokamak is prototype of X XI century reactor. Future International laboratory of thermonuclear materials testing and power engineering

    International Nuclear Information System (INIS)

    In 29-31 May of 2000 the presentation of the joint Kazakhstan-Russian draft of Kazakhstan material-testing tokamak (KTM) was carried out. KTM tokamak is implementing by decision of the President and Government of Republic of Kazakhstan for supporting of the Kazakhstan participation in development of draft within framework of ITER fusion reactor construction. Scientific head of the project is Russian academician - Velikhov V. and Russian Research Center 'Kurchatovskij Institute' , General designers - Scientific Research Institute for Electrophysical Equipment after D. V. Efremov (Russian Federation) and Kazakh Research Inst. for Energy Industry (KazNIIEhnergoprom). Scientific part of the project is working out in National Nuclear Center of Republic of Kazakhstan and Scientific Research Institute of Experimental and Theoretical Physics. KTM tokamak is experimental fusion device for materials testing study, as well as for designing of methods for protection of the reactor first wall, in-chamber elements and divertor planes, high frequency heat of antennas in energetic load regimes close to both the ITER and the future fusion energy reactors. KTM by it design presents spheric tokamak, which successfully combining advantages of the spheromaks (compactness) and the tokamaks (high plasma density). Now in the world there are similar operating spheric tokamaks: NSTX (USA), MAST (Great Britain), GLOBUS-M (Russian Federation). Principal peculiarity of KTM tokamak is existence of moving divertor device, which with help of manipulator allows to changing of examining samples without high vacuum disruption. Values of the thermal loads and fluences in the KTM are equal or higher than loads in operating tokamaks and correspond with ITER reactor loads. KTM tokamak will be the only mega-ampere device in the world with the aspect ratio A=2

  1. Electromagnetic microinstabilities in tokamak plasmas using a global spectral approach

    Energy Technology Data Exchange (ETDEWEB)

    Falchetto, G. L

    2002-03-01

    Electromagnetic microinstabilities in tokamak plasmas are studied by means of a linear global eigenvalue numerical code. The code is the electromagnetic extension of an existing electrostatic global gyrokinetic spectral toroidal code, called GLOGYSTO. Ion dynamics is described by the gyrokinetic equation, so that ion finite Larmor radius effects are taken into account to all orders. Non adiabatic electrons are included in the model, with passing particles described by the drift-kinetic equation and trapped particles through the bounce averaged drift-kinetic equation. A low frequency electromagnetic perturbation is applied to a low -but finite- {beta}plasma (where the parameter {beta} identifies the ratio of plasma pressure to magnetic pressure); thus, the parallel perturbations of the magnetic field are neglected. The system is closed by the quasi-neutrality equation and the parallel component of Ampere's law. The formulation is applied to a large aspect ratio toroidal configuration, with circular shifted surfaces. Such a simple configuration enables one to derive analytically the gyrocenter trajectories. The system is solved in Fourier space, taking advantage of a decomposition adapted to the toroidal geometry. The major contributions of this thesis are as follows. The electromagnetic effects on toroidal Ion Temperature Gradient driven (ITG) modes are studied. The stabilization of these modes with increasing {beta}, as predicted in previous work, is confirmed. The inclusion of trapped electron dynamics enables the study of its coupling to the ITG modes and of Trapped Electron Modes (TEM) .The effects of finite {beta} are considered together with those of different magnetic shear profiles and of the Shafranov shift. The threshold for the destabilization of an electromagnetic mode is identified. Moreover, the global formulation yields for the first time the radial structure of this so-called Alfvenic Ion Temperature Gradient (AITG) mode. The stability of the

  2. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form

  3. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  4. The disruptive instability in Tokamak plasmas

    NARCIS (Netherlands)

    Salzedas, F.J.B.

    2001-01-01

    Studies performed in RTP (Rijnhuizen Tokamak Project) of the most violent and dangerous instability in tokamak plasmas, the major disruption, are presented. A particular class of disruptions is analyzed, namely the density limit disruption, which occur in high density plasmas. The radiative te

  5. Physics of compact ignition tokamak designs

    International Nuclear Information System (INIS)

    Models for predicting plasma performance in compact ignition experiments are constructed on the basis of theoretical and empirical constraints and data from tokamak experiments. Emphasis is placed on finding transport and confinement models which reproduce results of both ohmically and auxiliary heated tokamak data. Illustrations of the application of the models to compact ignition designs are given

  6. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  7. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  8. Natural current profiles in a tokamak

    International Nuclear Information System (INIS)

    In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described

  9. Analysis of fast ion induced instabilities in tokamak plasmas

    CERN Document Server

    Horváth, László

    2015-01-01

    In magnetic confinement fusion devices like tokamaks, it is crucial to confine the high energy fusion-born helium nuclei ($\\alpha$-particles) to maintain the energy equilibrium of the plasma. However, energetic ions can excite various instabilities which can lead to their enhanced radial transport. Consequently, these instabilities may degrade the heating efficiency and they can also cause harmful power loads on the plasma-facing components of the device. Therefore, the understanding of these modes is a key issue regarding future burning plasma experiments. One of the main open questions concerning energetic particle (EP) driven instabilities is the non-linear evolution of the mode structure. In this thesis, I present my results on the investigation of $\\beta$-induced Alfv\\'{e}n eigenmodes (BAEs) and EP-driven geodesic acoustic modes (EGAMs) observed in the ramp-up phase of off-axis NBI heated plasmas in the ASDEX Upgrade tokamak. These modes were well visible on several line-of-sights (LOSs) of the soft X-ra...

  10. Magnet systems for ''Bean-Shaped'' tokamak

    International Nuclear Information System (INIS)

    Bean-shaping of tokamak plasmas offers a method of reaching stable operation at (beta) > 10%. In order to establish the indentation of the ''bean'', a set of high- current ''pushing coils'' (> 5 MA in a reactor) must be located at the midplane as close as possible to the inboard edge of the plasma. If located in the bore of the TF coils, then maintenance of the pushing coils may be impossible, and the interlocking coils may prevent reactor modularity. If located outside, the required pushing-coil current may be unacceptably large. This dilemma is overcome with a unique TF coil design in which the inboard leg is bent outward in the form of an arc. The pushing coils are housed in the midplane indentation of this arc, just outside the TF coils but adequately close to the plasma. The arched coil transfers forces to the top and bottom legs, where it can be reacted by a clamp structure if necessary. This technique would allow demountable joints to be placed near the inoard leg (for copper TF coils). Another design approach to the pushing coils is to use liquid Li or Na as the conductor and coolant. The liquid metal ''coils'' can be placed immediately adjacent to the plasma, giving optimal control of the plasma shape with minimal coil current, although modularity of the reactor may have to be surrendered. Conceptual designs are presented of PF and TF coil systems for an ignition test reactor with about 14% and for a full-scale demonstration reactor with about 20%, both using copper TF coils

  11. Instrumentation for plasma diagnosis in TN (Novillo Tokamak)

    International Nuclear Information System (INIS)

    In the Plasma Physics Laboratory of National Institute of Nuclear Research it has been utilized different devices for to determine electromagnetic parameters of Novillo Tokamak such as: magnetic fields, plasma currents, plasma column position and hoop voltage. For these measurements it was designed, constructed and calibrated magnetic soundings such as: magnetic field soundings, Rogowsky coil, coils of the type called sine/cosine and spires type riding saddle; as well as the electronic instrumentation associated with these devices. This electronics to be clear of instrumentation amplifiers for the detection of the soundings signals and differentiators utilized for the elimination of spurious induced currents in the soundings by the different Novillo electromagnetic fields. In this work is presented the methodology for the construction of this instruments, as well as the results of measurements effectuated in the two operation regimens of Tokamak: Cleaning discharge and Main discharge. (Author)

  12. Design of the welded bellows for KSTAR tokamak

    International Nuclear Information System (INIS)

    Vacuum vessel of the KSTAR(Korea Superconducting Tokamak Advanced Research) tokamak is a fully welded toroidal structure with noncircular cross-section nested in the TF(Toroidal Field) coil. According to the requirements of the physics design, sixteen horizontal ports, sixteen slanted ports, sixteen baking and cooling ports, and twenty-four top and bottom vertical ports are designed for the diagnostics, plasma heating, vacuum pumping, and baking and cooling. Bellows on these ports are used for flexible components to absorb the relative displacement due to the vacuum vessel thermal expansion and the electromagnetic force between the vacuum vessel and the cryostat ports. In this study, mechanical strength of the welded bellows for KSTAR vacuum vessel was evaluated

  13. Recent progress towards steady state tokamak operation with improved confinement in JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, Takaaki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2000-07-01

    In the JT-60U tokamak, optimization of high {beta}{sub p} mode and reversed shear mode plasmas are being done for establishment of scientific basis for steady state operation of tokamaks. In high {beta}{sub p} H-mode plasmas, {beta}{sub N}=2.9 and H{sub 89}=2.2-2.4 were sustained sanitarily by using high triangularity configuration and pressure profile optimization. Steady state performance was limited by resistive low toroidal mode number instabilities. Stabilization of resistive modes by using a newly installed ECRF system was attempted and a decrease of mode amplitude was observed but complete stabilization could not be achieved. In reversed shear plasmas, high fusion performance with equivalent DT fusion power gain of 0.5 was sustained for 0.8 s or an energy confinement time. The duration was limited by disruptive beta collapse that was encountered when the minimum value of q became 2 even with moderate beta, {beta}{sub N} {approx}1.2. Stationary sustainment of ITB was demonstrated in a full CD reversed shear plasma with LHCD. The sustainment of reversed shear current profile by bootstrap current was demonstrated in an ELMy H-mode edge reversed shear plasma with a high triangularity in a high q regime. A confinement enhancement factor of 3.5 and {beta}{sub N} of 2 were sustained for 2.7 s with stationary current and pressure profiles. Ar puffing to H mode plasmas aiming at high confinement with high density and high radiation fraction was performed and H{sub 89}{approx}1.4 with radiation fraction of 80% was obtained at 70% of Greenwald density. (author)

  14. Trade studies of plasma elongation for next-step tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Galambos, J.D.; Strickler, D.J.; Peng, Y.K.M.; Reid, R.L.

    1988-09-01

    The effect of elongation on minimum-cost devices is investigated for elongations ranging from 2 to 3. The analysis, carried out with the TETRA tokamak systems code, includes the effects of elongation on both physics (plasma beta limit) and engineering (poloidal field coil currents) issues. When ignition is required, the minimum cost occurs for elongations from 2.3 to 2.9, depending on the plasma energy confinement scaling used. Scalings that include favorable plasma current dependence and/or degradation with fusion power tend to have minimum cost at higher elongation (2.5-2.9); scalings that depend primarily on size result in lower elongation (/approximately/2.3) for minimum cost. For design concepts that include steady-state current-driven operation, minimum cost occurs at an elongation of 2.3. 12 refs., 13 figs.

  15. Charge exchange measurements of ion behavior in the ISX tokamak

    International Nuclear Information System (INIS)

    Analysis of charge exchange neutrals is the standard method for measuring ion behavior in tokamak plasmas. Limitations of this technique, which arise in dense and neutral-beam-heated plasmas, are discussed. Two refinements that address these limitations are described: a neutral particle analyzer, which incorporates both mass and energy resolution to distinguish different ion components such as the fast and thermal ions in beam-heated plasmas, and an improved data analysis procedure, which accounts for the effects of profiles and neutral attenuation in dense plasmas. A simple two-dimensional Monte Carlo neutral transport code was developed and used to investigate the effects of toroidally asymmetric neutral density profiles. Applications of these methods to experiments in the Impurity Study Experiment (ISX) are discussed; they include energy confinement studies on ISX-A, natural beam heating and high beta studies on ISX-B, and pellet fueling studies on ISX-B

  16. Effects of suprathermal fusion particles in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T.W.

    1978-01-01

    Several crucial properties of suprathermal (> 500 keV) fusion-products are explored, both in their initial phase and during their slowing-down period. A guiding center drift theory, which predicts the effect of energy loss on the motion of these suprathermals, is derived for a low-..beta.., symmetric (non-ripple) tokamak. Velocity-space scattering is ignored. Among the important implications of this theory are: (1) the net inward drift of fusion particles during their slow-down phase and (2) the importance of the plasma density and temperature in determining this drift. The effect the inward drifting has on the spatial profile for the suprathermals approaching thermal energies, on the energy distribution, and on the plasma heating profile is demonstrated for five reactor cases, ranging from near-term low-current devices to conceptual power reactors.

  17. Design and installation of a ferromagnetic wall in tokamak geometry.

    Science.gov (United States)

    Hughes, P E; Levesque, J P; Rivera, N; Mauel, M E; Navratil, G A

    2015-10-01

    Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak-Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics and overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability. PMID:26520952

  18. Ignition in a Tokamak using Deuterium - Helium 3

    International Nuclear Information System (INIS)

    Taking the characteristics of the ITER Tokamak project concerning the shape, the composition and the profile of the plasma, confinement time, beta and density limits, the size of a commercial D-T and D-3He reactor are compared, assuming loss by synchrotron radiation negligible. Reasonable size is obtained the D-3He reactor only with very high magnetic fields in the superconducting coil. With a realistic value of wall reflection coefficient for synchrotron radiation, plasma characteristics (purity, ash density, profiles should be improved for sizing a D-3He reactor. The case of pure deuterium reactor (i.e. without 3He injection) is also considered. Even with 3He contribution and neglecting losses by synchrotron radiation size is not realistic

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

    International Nuclear Information System (INIS)

    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

  20. Potential turbulence in tokamak plasmas

    International Nuclear Information System (INIS)

    Microscopic potential turbulence in tokamak plasmas are investigated by a multi-sample-volume heavy ion beam probe. The wavenumber/frequency spectra S(k,ω) of the plasmas potential fluctuation as well as density fluctuation are obtained for the first time. The instantaneous turbulence-driven particle flux, calculated from potential and density turbulence has oscillations of which amplitude is about 100 times larger than the steady-state outwards flux, showing sporadic behaviours. We also observed large-scale coherent potential oscillations with the frequency around 10-40 kHz. (author)

  1. The bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    The properties of the Hirshman equation for the bootstrap in the tokamak and the difference between it and the simpler Hinton-Hazeltine equation are discussed. The Hirshman model, which takes into account finite-aspect-ratio effects, is used to calculate the bootstrap current in the plasma in a circular cross section with Te = Ti. Approximate upper and lower bounds on the bootstrap current are obtained. These restrict the range of variation of the current as the temperature and density profiles vary. 16 refs., 9 figs

  2. Breakdown in the pretext tokamak

    International Nuclear Information System (INIS)

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges

  3. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  4. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    International Nuclear Information System (INIS)

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost

  5. Anomalous particle pinch in Tokamaks

    International Nuclear Information System (INIS)

    The diffusion coefficient in phase space usually varies with the particle energy. A consequence is the dependence of the fluid particle flux on the temperature gradient. If the diffusion coefficient in phase space decreases with the energy in the bulk of the thermal distribution function, the particle thermodiffusion coefficient which links the particle flux to the temperature gradient is negative. This is a possible explanation for the inward particle pinch that is observed in tokamaks. A quasilinear theory shows that such a thermodiffusion is generic for a tokamak electrostatic turbulence at low frequency. This effect adds to the particle flux associated with the radial gradient of magnetic field. This behavior is illustrated with a perturbed electric potential, for which the trajectories of charged particle guiding centers are calculated. The diffusion coefficient of particles is computed and compared to the quasilinear theory, which predicts a divergence at low velocity. It is shown that at low velocity, the actual diffusion coefficient increases, but remains lower than the quasilinear value. Nevertheless, this differential diffusion between cold and fast particles leads to an inward flux of particles. (author)

  6. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime

  7. Cluster storage for COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pisacka, J., E-mail: pisacka@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Hron, M., E-mail: hron@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University, V Holesovickach 2, 180 00 Praha 8 (Czech Republic); Panek, R., E-mail: panek@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer New data storage system needed for the COMPASS tokamak. Black-Right-Pointing-Pointer Distributed, fault-tolerant, parallel, scalable, non-proprietary. Black-Right-Pointing-Pointer GlusterFS selected for testing on a small test bed. Black-Right-Pointing-Pointer Aggregated reading throughput reached 300 MiB/s for 6 clients - very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  8. Cluster storage for COMPASS tokamak

    International Nuclear Information System (INIS)

    Highlights: ► New data storage system needed for the COMPASS tokamak. ► Distributed, fault-tolerant, parallel, scalable, non-proprietary. ► GlusterFS selected for testing on a small test bed. ► Aggregated reading throughput reached 300 MiB/s for 6 clients – very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  9. Predictive Modeling of Tokamak Configurations*

    Science.gov (United States)

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

  10. Tokamak Physics Experiment divertor design

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m2. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services

  11. Atomic physics in tokamak plasmas

    International Nuclear Information System (INIS)

    Tokamak discharges produce hydrogen-isotope plasmas in a quasi-steady state, with radial electron temperature, Tsub(e)(r), and density nsub(e)(r), distribution usually centrally peaked, with typical values Tsub(e)(0) approx.= 1 - 3 keV, nsub(e)(r) approx.= 1014 cm-3. Besides hydrogen, the plasma contains small quantities of carbon, oxygen, various construction or wall-conditioning materials such as Fe, Cr, Ni, Ti, Zr, Mo, and perhaps elements added for special diagnostic purposes, e.g., Si, Sc, Al, or noble gases. These elements are spatially fairly homogeneously distributed, with the different ionization states occurring near radial locations where Tsub(e)(r) approx.= Esub(i), the ionization potential. Thus, spectroscopic measurements of various plasma properties, such as ion temperatures, plasma motions or oscillations, radial transport rates, etc. are automatically endowed with spatial resolution. Furthermore the emitted spectra, even of heavier elements such as Fe or Ni, are fairly simple because only the ground levels are appreciably populated under the prevailing plasma conditions. Identification of near-ground transitions, including particularly magnetic dipole and intercombination transitions of ions with ionization potentials in the several keV range, and determination of their collisional and radiative transition probabilities will be required for development of appropriate diagnostics of tokamak-type plasma approaching the prospective fusion reactor conditions. (orig.)

  12. Tokamak equilibria with nearly zero central current: the current hole

    International Nuclear Information System (INIS)

    The observation of stable sustainment of the 'current hole', namely the nearly zero current density region in the central part of a tokamak plasma, has opened a new class of configurations in tokamak plasmas, and a variety of research from the viewpoints of equilibrium, magnetohydrodynamics (MHD) stability, particle orbits and radial transport has been generated. Some theories and codes have been tested and extended by being applied to extreme conditions in the current hole with very weak poloidal field. The current hole is generated due to a transient negative toroidal electric field established when a large off-axis non-inductive current is rapidly formed. It has been observed in high confinement plasmas with a large fraction of bootstrap current in advanced tokamak operation. The current hole is very stiff against current drive, which suggests that it is a saturated or self-organized system. Appearance of the current hole in ITER and DEMO would be expected in some of the operation scenarios, and its influence and its control methods have been studied. Results of experimental and theoretical studies on the current hole are reviewed. (review article)

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

    International Nuclear Information System (INIS)

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

  14. Analysis of neutral hydrogenic emission spectra in a tokamak

    International Nuclear Information System (INIS)

    Balmer-α radiation by the excitation of thermal and fast neutral hydrogenic particles has been investigated in a magnetically confined fusion device, or tokamak, from the Korea Superconducting Tokamak Advanced Research (KSTAR). From the diagnostic point of view, the emission from thermal neutrals is associated with passive spectroscopy and that from energetic neutrals that are usually injected from the outside of the tokamak to the active spectroscopy. The passive spectroscopic measurement for the thermal Balmer-α emission from deuterium and hydrogen estimates the relative concentration of hydrogen in a deuterium-fueled plasma and therefore, makes a useful tool to monitor the vacuum wall condition. The ratio of hydrogen to deuterium obtained from this measurement qualitatively correlates with the energy confinement of the plasma. The Doppler-shifted Balmer-α components from the fast neutrals features the spectrum of the motional Stark effect (MSE) which is an essential principle for the measurement of the magnetic pitch angle profile. Characterization of this active MSE spectra, especially with multiple neutral beam lines crossing along the observation line of sight, has been done for the guideline of the multi-ion-source heating beam operation and for the optimization of the narrow bandpass filters that are required for the polarimeter-based MSE diagnostic system under construction at KSTAR

  15. Synergistic effects of magnetic and velocity shear on electromagnetic drift modes in tokamaks

    International Nuclear Information System (INIS)

    Full text: In the fluid model, electromagnetic drift mode in low beta tokamak plasmas is studied from a set of new-derived eigenequations, including magnetic shear, perpendicular and parallel velocity shears, perpendicular and parallel current density and their shears, and finite beta. It is found that there exists a threshold of perpendicular velocity shear, at which the growth rate tends to zero. The threshold increases with decrease of magnetic shear. On the other hand, the increase of beta reduces the growth rate but increases the velocity shear threshold. In addition, we study the effects of parallel velocity shear on the instability and find that it enhances the instability. Furthermore, the preliminary calculations show that the perpendicular current density shear suppresses the instability while the parallel current density shear strengthens it. (author)

  16. Interactive exploration of tokamak turbulence simulations in virtual reality

    International Nuclear Information System (INIS)

    We have developed an immersive visualization system designed for interactive data exploration as an integral part of our computing environment for studying tokamak turbulence. This system of codes can reproduce the results of simulations visually for scrutiny in real time, interactively and with more realism than ever before. At peak performance, the VR system can present for view some 400 coordinated images per second. The long term vision this approach targets is a open-quote holodeck-like close-quote virtual-reality environment in which one can explore gyrofluid or gyrokinetic plasma simulations interactively and in real time, visually, with concurrent simulations of experimental diagnostic devices. In principle, such a open-quote virtual tokamak close-quote computed environment could be as all encompassing or as focussed as one likes, in terms of the physics involved. The computing framework in one within which a group of researchers can work together to produce a real and identifiable product with easy access to all contributions. This could be our version of NASA's next generation Numerical Wind Tunnel. The principal purpose of this VR capability for Numerical Tokamak simulation is to provide interactive visual experience to help create new ways of understanding aspects of the convective transport processes operating in tokamak fusion experiments. The effectiveness of the visualization method is strongly dependent on the density of frame-to-frame correlation. Below a threshold of this quantity, short term visual memory does not bridge the gap between frames well enough for there to exist a strong visual connection. Above the threshold, evolving structures appear clearly. The visualizations show the 3D structure of vortex evolution and the gyrofluid motion associated with it. We discovered that it was very helpful for visualizing the cross field flows to compress the virtual world in the toroidal angle

  17. Project of magnetic fluctuation measurement by cross polarization scattering in the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Up to now the direct measurement of the magnetic fluctuations in the core of tokamak plasmas has been impossible. According to various turbulence models, the estimated level of B/B lies between 10-6 (residual magnetic fluctuations due to drift wave i.e. O(βn/n) where β is the toroidal beta) and 10-4 (obtained assuming that the heat transport is only due to magnetic turbulence). We describe in this paper a tentative diagnostic to measure these fluctuations. It is based on the cross polarization scattering: Magnetic fluctuations scatter an incident electromagnetic wave with a change in polarization. In section II the basic principle of the diagnostic is explained. In section III the experimental set up is described. It will be tested on the Tore Supra Tokamak at the end of 1990. (author) 2 refs., 2 figs

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

    Indian Academy of Sciences (India)

    D Raju; R Jha; P K Kaw; S K Mattoo; Y C Saxena; Aditya Team

    2000-11-01

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as the discharge progresses. It is observed that during the current rise phase, current perturbation undergoes transition from = 5 poloidal structure to = 4 and then to = 3. At the time of current termination, = 2 perturbation is observed. It is observed that the mode frequency remains nearly constant (≈10 kHz) when poloidal mode structure changes from = 4 to = 2. This may be either an indication of mode coupling or a consequences of changes in the plasma electron temperature and density scale length.

  19. Physics Basis and Simulation of Burning Plasma Physics for the Fusion Ignition Research Experiment (FIRE)

    International Nuclear Information System (INIS)

    The FIRE [Fusion Ignition Research Experiment] design for a burning plasma experiment is described in terms of its physics basis and engineering features. Systems analysis indicates that the device has a wide operating space to accomplish its mission, both for the ELMing H-mode reference and the high bootstrap current/high beta advanced tokamak regimes. Simulations with 1.5D transport codes reported here both confirm and constrain the systems projections. Experimental and theoretical results are used to establish the basis for successful burning plasma experiments in FIRE

  20. Interlock system for the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Hron, Martin; Sova, J.; Šíba, J.; Kovář, J.; Adámek, Jiří; Pánek, Radomír; Havlíček, Josef; Písačka, Jan; Mlynář, Jan; Stöckel, Jan

    2010-01-01

    Roč. 85, 3-4 (2010), s. 505-508. ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition and Remote Participation for Fusion Research/7th./. Aix – en – Provence, 15.06.2009-19.06.2009] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak operation * Interlock * Personnel safety Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.143, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V3C-5003BXW-1&_user=6542793&_coverDate=07%2F31%2F2010&_rdoc=1&_fmt=high&_orig=search&_origin=search&_sort=d&_docanchor=&view=c&_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=ef5794d05cc6530a905d1de43aa0ac6a&searchtype=a

  1. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

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

  2. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

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

  3. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  4. Synchrotron radiation in inhomogeneous tokamak plasmas

    International Nuclear Information System (INIS)

    Synchrotron emission in a tokamak configuration with inhomogeneous plasma parameters is considered to investigate the effects of the temperature profile and vertical elongation on the radiation loss. Using the numerical solution of the transfer equation for ITER-like plasma parameters, several new results on the radiated energy in a Maxwellian plasma have been derived. In particular: (i) synchrotron loss is profile dependent, namely, at constant average thermal energy, the emitted radiation increases with the peak temperature, (ii) an analytical formula of the global loss in inhomogeneous tokamak plasmas with arbitrary vertical elongation is established, (iii) the maximum of the frequency emission spectrum is a linear function of the volume average temperature, (iiii) high frequency synchrotron radiation is entirely due to electrons with energy much greater than the thermal energy. The need for experimental investigations on synchrotron emission in present-day large tokamaks to determine the effect of reflections of the complex tokamak first wall is stressed

  5. Overview of JT-60U progress towards steady-state advanced tokamak

    International Nuclear Information System (INIS)

    Recent experimental results on steady state advanced tokamak (AT) research on JT-60U are presented with emphasis on longer time scale in comparison with characteristics time scales in plasmas. Towards this, modification on control in operation, heating and diagnostics systems have been done. As the results, ∼ 60 s Ip flat top and an ∼ 30 s H-mode are obtained. The long pulse modification has opened a door into a new domain for JT-60U. The high normalized beta (βN) of 2.3 is maintained for 22.3 s and 2.5 for 16.5 s in a high βp H-mode plasma. A standard ELMy H-mode plasma is also extended and change in wall recycling in such a longer time scale has been unveiled. Development and investigation of plasmas relevant to AT operation has been continued in former 15 s discharges as well in which higherNB power (≤ 10 s) is available. Higher βN ∼ 3 is maintained for 6.2 s in high βp H-mode plasmas. High bootstrap current fraction (fBS) of ∼ 75% is sustained for 7.4 s in an RS plasma. On NTM suppression by localized ECCD, ECRF injection preceding the mode saturation is found to be more effective to suppress the mode with less power compared to the injection after the mode saturated. The domain of the NTM suppression experiments is extended to the high βN regime, and effectiveness of m/n=3/2 mode suppression by ECCD is demonstrated at βN ∼ 2.5-3. Genuine center-solenoid less tokamak plasma start up is demonstrated. In a current hole region, it is shown that no scheme drives a current in any direction. Detailed measurement in both spatial and energy spaces of energetic ions showed dynamic change in the energetic ion profile at collective instabilities. Impact of toroidal plasma rotation on ELM behaviors is clarified in grassy ELM and QH domains. (author)

  6. D-D tokamak reactor studies

    International Nuclear Information System (INIS)

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated

  7. Plasma diagnostics using synchrotron radiation in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  8. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aBtx of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

  9. Epoxide insulation for Tokamak coils

    International Nuclear Information System (INIS)

    The construction and testing of 12-tonne toroidal-field electromagnets for the Joint European Torus by Brown Boveri and Cie (Mannheim) are described. The principle of Tokamak confinement of a plasma which acts as the secondary winding of a transformer is explained. The Cu conductors are sanded and coated with epoxide adhesive before being wrapped in 7mm thick woven glass fibre, dried by heating under vacuum, impregnated and encapsulated in 1.2 tonnes of Araldite, which is solidified under pressure of 4 atmospheres and hardened for ten hours at 1500C. The prototype withstood tests involving 25,000 flexure cycles at 1.1 MN and 2 Hz, 2,000 quarter-hour 10kA heating cycles between 840 and 200C, and exposure to 500 million rads. 32 such coils were constructed at the rate of one every three weeks. (M.B.D.)

  10. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict

  11. Simultaneous evolution of plasma rotation, radial electric field, MHD activity and plasma confinement in the STOR-M tokamak

    International Nuclear Information System (INIS)

    Radial electric field shear and poloidal plasma rotation are important factors affecting transport and confinement in tokamaks. Alteration of the electric field and plasma rotation in the vicinity of magnetic islands is also an important factor in tokamak plasma confinement. In the STOR-M tokamak, fast (∼1 ms) simultaneous alterations of the radial electric field, plasma rotation (Mparallel = 0-0.4 in the plasma current direction), floating potential fluctuations in the periphery and MHD activity generated by rotating islands have been observed experimentally during normal ohmic discharges. The observed time and magnitude of the changes depend on the average electron density and poloidal beta at the beginning of the discharge. In discharges with high initial poloidal beta these changes are accompanied by a reduction in Hα emission and an increase in the line averaged density. Drastic decreases in Hα and increases in line averaged electron density and estimation of poloidal beta suggest that STOR-M confinement is significantly affected in ohmic discharges without an external additional energy input or biasing. MHD activity in STOR-M is damped when a negative electric field is observed at the limiter region of the plasma edge. MHD frequency is observed to decrease with the negative electric field

  12. PROGRESS TOWARD SUSTAINED HIGH-PERFORMANCE ADVANCED TOKAMAK DISCHARGES IN DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    J.R. FERRON; D.P. BREEAN; T.A. CASPER; A.M. GAROFALO; C.M. GREENFIELD; A.W. HYATT; R.J. JAYAKUMAR; L.C. JOHNSON; J.E. KINSEY; R.J. LaHAYE; L.L. LAO; E.A. LAZARUS; J. LOHR; T.C. LUCE; M. MURAKAMI; M. OKABAYASHI; C.C. PETTY; P.A. POLITZER; R. PRATER; H. REIMERDES; E.J. STRAIT; T.S. TAYLOR; A.D. TURNBULL; J.G. WATKINS; M.R. WADE; W.P. WEST

    2002-07-01

    Key elements of a sustained advanced tokamak discharge in DIII-D are a large fraction of the total current from bootstrap current (f{sub BS}) and parameters that optimize the capability to use electron cyclotron current drive (ECCD) at {rho} {approx} 0.5 to maintain the desired current profile [1-4]. Increased f{sub BS} results from increasing both the normalized beta ({beta}{sub N}) and the minimum value of the safety factor (q{sub min}). Off-axis ECCD is, for the available gyrotron power, optimized at high {beta}{sub N}, high electron temperature (T{sub e}) and low electron density (n{sub e}). As previously reported [2-4], these required elements have been separately demonstrated: density control at high {beta}{sub N} with n{sub e} {le} 5 x 10{sup 19} m{sup -3} using divertor-region pumping, stability at high {beta}, and off-axis ECCD at the theoretically predicted efficiency. This report summarizes recent work on optimizing and integrating these results through evaluation of the dependence of the beta limit on q{sub min} and q{sub 95}, exploration of discharges with relatively high q{sub min}, testing of feedback control of T{sub e} for control of the q profile evolution, and modification of the current profile time evolution when ECCD is applied.

  13. Potential commercial reactor based on a small tokamak hybrid design

    International Nuclear Information System (INIS)

    An ignition tokamak reactor design has been obtained which represents a starting point for the conceptual design of a 1000 MW/sub e/ commercial system. The design includes Nb3Sn superconducting coils (TF, OH, and SF), water-cooled fissile blanket (e.g., uranium oxide), positive-ion based neutral beams with no direct energy recovery, and an ignited plasma with a = 0.9 m and an aspect ratio A = 4.0. The TF coil bore has a vertical bore of 7 m and a horizontal bore of 5 m both of which are a factor of two larger than the corresponding bore dimensions of the LCP (Large Coil Project) TF coil. The plasma is characterized as follows: stability factor q = 2.5, Z/sub eff/ approx. 1, poloidal beta β/sub p/ less than or equal to A, a elongation delta in the range between 1.6 and 1.7. A number of potential operating conditions for the plasma and device have been identified for which the plasma beta β lies within the range from 6.5% to 7.3%, and the plasma temperature has an average value between 11 keV and 12.5 keV. The design was obtained using the computer code COAST and represents a self-consistent sizing and costing result

  14. BETA-S, Multi-Group Beta-Ray Spectra

    International Nuclear Information System (INIS)

    1 - Description of program or function: BETA-S calculates beta-decay source terms and energy spectra in multigroup format for time-dependent radionuclide inventories of actinides, fission products, and activation products. Multigroup spectra may be calculated in any arbitrary energy-group structure. The code also calculates the total beta energy release rate from the sum of the average beta-ray energies as determined from the spectral distributions. BETA-S also provides users with an option to determine principal beta-decaying radionuclides contributing to each energy group. The CCC-545/SCALE 4.3 (or SCALE4.2) code system must be installed on the computer before installing BETA-S, which requires the SCALE subroutine library and nuclide-inventory generation from the ORIGEN-S code. 2 - Methods:Well-established models for beta-energy distributions are used to explicitly represent allowed, and 1., 2. - and 3. -forbidden transition types. Forbidden non-unique transitions are assumed to have a spectral shape of allowed transitions. The multigroup energy spectra are calculated by numerically integrating the energy distribution functions using an adaptive Simpson's Rule algorithm. Nuclide inventories are obtained from a binary interface produced by the ORIGEN-S code. BETA-S calculates the spectra for all isotopes on the binary interface that have associated beta-decay transition data in the ENSDF-95 library, developed for the BETA-S code. This library was generated from ENSDF data and contains 715 materials, representing approximately 8500 individual beta transition branches. 3 - Restrictions on the complexity of the problem: The algorithms do not treat positron decay transitions or internal conversion electrons. The neglect of positron transitions in inconsequential for most applications involving aggregate fission products, since most of the decay modes are via electrons. The neglect of internal conversion electrons may impact on the accuracy of the spectrum in the low

  15. Non-inductive current drive by EC waves in an inboard poloidal magnetic field null configuration on the spherical tokamak QUEST

    International Nuclear Information System (INIS)

    To investigate the effects of confinement of fast electrons on poloidal beta equilibrium, experiments under different toroidal magnetic mirror ratios, MOMFC, were performed in electron cyclotron heated plasma in the spherical tokamak QUEST. The proportional constants of the relation between plasma current and vertical magnetic fields decreased with increase in MOMFC. The poloidal field null point appeared in the chamber for MOMFC > 1.8 at a plasma current value of ∼15 kA. Poloidal beta and temperature of hard X-rays increased with MOMFC. This suggests that high poloidal beta equilibrium plasma is produced by better confinement of energetic trapped particles. (author)

  16. High harmonic fast waves in high beta plasmas

    International Nuclear Information System (INIS)

    High harmonic fast magnetosonic wave in high beta/high dielectric plasmas is investigated. including the finite-Larmor-radius effects. In this regime, due to the combination of group velocity slow down and the high beta enhancement, the electron absorption via electron Landau and electron magnetic pumping becomes significant enough that one can expect a strong (∼ 100%) single pass absorption. By controlling the wave spectrum, the prospect of some localized electron heating and current drive appears to be feasible in high beta low-aspect-ratio tokamak regimes. Inclusion of finite-Larmor-radius terms shows an accessibility limit in the high ion beta regime (βi = 50% for a deuterium plasma) due to mode-conversion into an ion Bernstein-wave-like mode while no beta limit is expected for electrons. With increasing ion beta, the ion damping can increase significantly particularly near the beta limits. The presence of energetic ion component expected during intense NBI and α-heating does not appear to modify the accessibility condition nor cause excessive wave absorption

  17. Study of electron density and its fluctuations in tokamaks plasmas by fast infrared interferometry

    International Nuclear Information System (INIS)

    The electron density knowledge in tokamak plasma is fundamental for controlled fusion research. Its study can be made by interferometric measurement of plasma refraction index. Density and density fluctuation measurements are given for present and future tokamak, the wavelength used must be in the far infrared. The interferometer used type employs two identical lasers. Waveguide type submillimetric lasers, optically pumped by a CO2 laser, have been developed and optimized. Detectors used are Schottky diodes. The interferometer allows a radial study of the plasma and presents a great stability during the measurement

  18. Applications of non-resonant RF forces for improvement of tokamak reactor performances, 1

    International Nuclear Information System (INIS)

    Applications of Radio Frequency to a tokamak divertor plasma for improvement of its reactor relevancy is studied and proposed. RF is applied to a divertor region of a tokamak by use of wave guide launchers on consideration of a reactor environment. Since the ponderomotive force is dependent on the charge to mass ratio of ions, various useful applications are considered. They covers some of the recent key issues in the development research toward nuclear fusion: the reduction of the heat load on the divertor plate, improvement of tritium inventory, impurity control, and helium ash removal. (author)

  19. Importance of the fine structure in a tokamak for the abnormal transport and the internal disruptions

    International Nuclear Information System (INIS)

    The problem of energy transport in a Tokamak, in presence of magnetic islets, has been treated by decomposing this problem in different bricks. To assembly the different bricks the model of dynamic percolation, which couples by the intermediate of scattering coefficient, the activity of transport sites (islets size) to the profile of transported quantity (temperature profile) has been chosen. The results, got with this model, results connected to the hypothesis of a limited number of islets, agree with the different observations. A possible application of this model could be the exploration of different operating conditions of Tokamak and a research of improved confinement running. (N.C.). 149 refs., 85 figs

  20. Annual progress report on fusion plasma theory task III: auxiliary heating in tokamaks and tandem mirrors

    International Nuclear Information System (INIS)

    The research we have accomplished during the past year has focussed on ICRF coupling, heating and breakeven studies for tokamaks and ECRF fundamental second harmonic heating in tandem mirrors. The studies have included ICRF Fokker-Planck heating and breakeven studies for large tokamaks such as JET, fundamental work on a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas, a formulation and code development for ICRF waveguide coupling in tokamak edge regions. ECRF ray tracing studies have been carried out for fundamental and second harmonic propagation, absorption and whistler microinstabilities in tandem mirror plug and barrier regions of Phaedrus, TMX-U and TASKA. The two-dimensional velocity space, time dependent Fokker-Planck heating studies have concentrated on D-T breakeven scenarios for fundamental minority deuterium and second harmonic tritium regimes

  1. On Use of Semiconductor Detector Arrays on COMPASS Tokamak

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Imríšek, Martin; Havlíček, Josef; Mlynář, Jan; Naydenkova, Diana; Háček, Pavel; Hron, Martin; Janky, Filip; Sarychev, D.; Berta, M.; Bencze, A.; Szabolics, T.

    -, č. 71 (2012), s. 844-850. ISSN 2010-376X. [ICPP 2012 : International Conference on Plasma Physics. Venice, 14.11.2012-16.11.2012] R&D Projects: GA ČR GA202/09/1467; GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : bolometry * plasma diagnostics * soft X-rays * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics https://www.waset.org/journals/waset/v71/v71-143.pdf

  2. A DESIGN RETROSPECTIVE OF THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    OAK-B135 The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and rf heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research program. This paper gives an integrated picture of the facility and its capabilities

  3. A design retrospective of the DIII-D tokamak

    Science.gov (United States)

    Luxon, J. L.

    2002-05-01

    The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and RF heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research programme. An integrated picture of the facility and its capabilities is presented.

  4. EBW H&CD Potential for Spherical Tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    Vol. 1406. New York : American Institute of Physics, 2011 - (Phillips, C.; Wilson, J.), s. 477-480 ISBN 978-0-7354-0978-1. - (AIP Conference Proceedings. 1406). [Topical Conference on Radio Frequency Power in Plasmas/19./. Newport (US), 01.06.2011-03.06.2011] R&D Projects: GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : Fusion * tokamak * heating * current drive * electron Bernstein wave * EBW Subject RIV: BL - Plasma and Gas Discharge Physics http://scitation.aip.org/getpdf/servlet/GetPDFServlet?filetype=pdf&id=APCPCS001406000001000477000001&idtype=cvips&doi=10.1063/1.3665018&prog=normal

  5. Vacuum Control and Gas Handling for COMPASS Tokamak

    Czech Academy of Sciences Publication Activity Database

    Janky, F.; Pereira, T.V.; Santos, B.A.; Hron, Martin

    Vol. 2. Prague: MATFYZPRESS, Prague, 2009 - (Šafránková, J.; Pavlů, J.), s. 153-157 ISBN 978-80-7378-102-6. [Annual conference of doctoral students - WDS 2009 /18./. Prague (CZ), 02.06.2009-05.06.2009] R&D Projects: GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : vacuum * control * tokamak COMPASS * gas handling Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/pdf09/WDS09_226_f2_Janky.pdf

  6. Computer and engineering calculations of Brazilian Tokamak-II

    International Nuclear Information System (INIS)

    Analytical and computer calculations carried out by researches of Physics Institute - University of Sao Paulo (IFUSP), for defining the engineering project and constructing the TBR-II tokamak are presented. The hydrodynamics behavioue and determined parameters for magnetic confinement of the plasma were analysed. The computer code was developed using magnetohydrodynamics (MHD) equations which involve plasma interactions, magnetic field and electrical current circulating in more than 20 coils distributed around toroidal vase of the plasma. The electromagnetic, thermal and mechanical couplings are also presented. The TBR-II will be feed by two turbo-generators with 15 MW each one. (M.C.K.)

  7. First dedicated observations of runaway electrons in the COMPASS tokamak

    Directory of Open Access Journals (Sweden)

    Vlainić Miloš

    2015-06-01

    Full Text Available Runaway electrons present an important part of the present efforts in nuclear fusion research with respect to the potential damage of the in-vessel components. The COMPASS tokamak a suitable tool for the studies of runaway electrons, due to its relatively low vacuum safety constraints, high experimental flexibility and the possibility of reaching the H-mode D-shaped plasmas. In this work, results from the first experimental COMPASS campaign dedicated to runaway electrons are presented and discussed in preliminary way. In particular, the first observation of synchrotron radiation and rather interesting raw magnetic data are shown.

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

    Czech Academy of Sciences Publication Activity Database

    Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud

    2012-01-01

    Roč. 356, č. 1 (2012), s. 012007. ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf

  9. Motion of Charged Particles in Perturbed Magnetic Fields of Tokamak

    Czech Academy of Sciences Publication Activity Database

    Papřok, R.; Krlín, Ladislav; Cahyna, Pavel; Riccardo, V.

    Vol. 2. Prague: MATFYZPRESS, Prague, 2009 - (Šafránková, J.; Pavlů, J.), s. 139-143 ISBN 978-80-7378-102-6. [Annual conference of doctoral students - WDS 2009 /18./. Prague (CZ), 02.06.2009-05.06.2009] Institutional research plan: CEZ:AV0Z20430508 Keywords : resonant magnetic perturbations * magnetic islands * electric field * runaway electrons * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/pdf09/WDS09_223_f2_Paprok.pdf

  10. First measurements with U-probe on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Kovařík, Karel; Ďuran, Ivan; Stöckel, Jan; Seidl, Jakub; Šesták, David; Brotánková, J.; Spolaore, M.; Martines, E.; Vianello, N.

    Vol. 2. Prague : MATFYZPRESS, 2013 - (Šafránková, J.; Pavlů, J.), s. 109-114 ISBN 978-80-7378-251-1. - (WDS). [Annual Conference of Doctoral Students – WDS 2013 /22./. Praha (CZ), 04.06.2013-07.06.2013] R&D Projects: GA MŠk 7G09042; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/proc/pdf13/WDS13_217_f2_Kovarik.pdf

  11. Focus on nuclear fusion research

    Czech Academy of Sciences Publication Activity Database

    Křenek, Petr; Mlynář, Jan

    2011-01-01

    Roč. 61, - (2011), s. 62-63. ISSN 0375-8842 Institutional research plan: CEZ:AV0Z20430508 Keywords : ITER * COMPASS * fusion energy * tokamak * EURATOM Subject RIV: BL - Plasma and Gas Discharge Physics http://www.ipp.cas.cz/Tokamak/clanky/energetika_COMPASS.pdf

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

    International Nuclear Information System (INIS)

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

  13. Ideal and resistive MHD stability of internal kink modes in circular and shaped tokamaks

    International Nuclear Information System (INIS)

    Recent results for the MHD stability of internal kink modes in tokamaks are reviewed. In general, ideal stability is more restrictive than the conventionally cited limit βpp is the poloidal beta at the q = 1 surface). This holds, in particular, for shaped equilibria, where low shear in combination with elliptic shaping can drastically reduce the pressure limit. Also in resistive MHD, interchange effects are frequently destabilizing, and resistive stability at βp≥0.05 is achieved, for circular section, only with a very restricted class of current profiles, and not at all for JET-shaped cross section. (author) 9 figs., 24 refs

  14. Advanced tokamak physics experiments on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, T.S. [General Atomics, San Diego, CA (United States)

    1998-12-01

    Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy confinement time ({tau}{sub E}) and the plasma pressure (or beta {beta}{sub T} = 2 {mu}{sub 0} < p > /B{sub T}{sup 2}) can be achieved in steady-state conditions with high self driven bootstrap current fraction. In addition, effective power exhaust and impurity and particle control is required. Significant progress has been made in experimentally achieving regimes having the required performance in all of these aspects as well as in developing a theoretical understanding of the underlying physics. The authors have extended the duration of high performance ELMing H-mode plasmas with {beta}{sub N} H{sub iop} {approximately} 10 for 5 {tau}{sub E} ({approximately}1 s) and have demonstrated that core transport barriers can be sustained for the entire 5-s neutral beam duration in L-mode plasmas. Recent DIII-D work has advanced the understanding of improved confinement and internal transport barriers in terms of E x B shear stabilization of micro turbulence. With the aim of current profile control in discharges with negative central magnetic shear, they have demonstrated off-axis electron cyclotron current drive for the first time in a tokamak, finding an efficiency above theoretical expectations. MHD stability has been improved through shape optimization, wall stabilization, and modification of the pressure and current density profiles. Heat flux reduction and improved impurity and particle control have been realized through edge/divertor radiation and understanding and utilization of forced scrape off layer flow and divertor baffling.

  15. Particle and energy balances in tokamak plasmas

    International Nuclear Information System (INIS)

    Computational and experimental studies on particle and energy balances in tokamak plasmas are described. Firstly, concerning the modeling of tokamak plasmas, the particle balance considering diffusion and recycling, and the energy balance considering transport and energy losses due to impurities are discussed. Production mechanisms of gaseous and metallic impurities, which play important role in tokamak plasmas, are also discussed from a viewpoint of plasma-wall interactions. Scaling laws of density, temperature and energy confinement time are shown on the basis of recent data. Secondarily, tokamak plasmas are simulated with the above model, and anomalous diffusion and electron thermal conduction are indicated. Characteristics of a future tokamak plasma are also simulated. Stationary impurity density distributions and related energy losses, such as bremsstrahlung, ionization and excitation, are calculated taking into account diffusion and ionization processes. Edge cooling by oxygen impurities is described quantitatively compared with experiments. Permissible impurity levels of carbon, oxygen and iron in future large tokamaks are estimated. Thirdly, experimental studies on surface cleaning methods of the first wall are described; discharge cleaning in JFT-2, baking effect on the outgassing rates of wall materials, surface treatment of high-temperature molybdenum by oxygen and hydrogen gases, and in-situ coating of molybdenum by a coaxial magnetron sputter method. Lastly, problems in future large tokamaks aiming at break-even or self-ignited plasma are discussed quantitatively, such as trapped particle instabilities, impurities and additional heating. It is predicted that new conceptions will be necessary to overcome the problems and attain the fusion goal. (auth.)

  16. Short-term power sources for tokamaks and other physical experiments

    Czech Academy of Sciences Publication Activity Database

    Zajac, Jaromír; Žáček, František; Brettschneider, Zbyněk; Lejsek, V.

    2007-01-01

    Roč. 82, č. 4 (2007), s. 369-379. ISSN 0920-3796 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * Impulse power sources * Energy accumulation Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 1.058, year: 2007 http://www.sciencedirect.com/science/journal/09203796

  17. 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

  18. Emissive probe measurements of plasma potential fluctuations in the edge plasma regions of tokamaks

    Czech Academy of Sciences Publication Activity Database

    Balan, P.; Schrittweiser, R.; Ionita, C.; Cabral, J. A.; Figueiredo, H. F. C.; Fernandes, H.; Varandas, C.; Adámek, Jiří; Hron, Martin; Stöckel, Jan; Martines, E.; Tichý, M.; Van Oost, G.

    2003-01-01

    Roč. 74, č. 3 (2003), s. 1583-1587. ISSN 0034-6748 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : plasma physics, tokamaks, probes Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.343, year: 2003

  19. Measurement of the Fluctuation-Induced Flux with Emissive Probe in the CASTOR Tokamak

    Czech Academy of Sciences Publication Activity Database

    Balan, P.; Adámek, Jiří; Ďuran, Ivan; Hron, Martin; Ionita, C.; Martines, E.; Schrittwieser, R.; Stöckel, Jan; Tichý, M.; Van Oost, G.

    Mulhouse: European Physical Society, 2002 - (Behn, R.; Varandas, C.), s. P-2.072 [EPS Conference on Controlled Fusion and Plasma Physics /29./. Montreux (CH), 17.06.2002-21.06.2002] Institutional research plan: CEZ:AV0Z2043910 Keywords : CASTOR tokamak, electric field Subject RIV: BL - Plasma and Gas Discharge Physics

  20. 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

  1. Engineering overview of the National Spherical Tokamak Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chrzanowski, J.H.; Fan, H.M.; Heitzenroeder, P.J.; Ono, M.; Robinson, J. [Princeton Plasma Physics Lab., NJ (United States)

    1995-12-31

    The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device designed for a plasma current of 1 MA. It features auxiliary heating and current drive and a close-fitting conducting shell to maximize plasma pressure. NSTX is designed for a 5 sec. experimental pulse to demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The device will be sited in the former Princeton Large Torus (PLT) test cell and will utilize the PLT radiation shielding, base structure, and cell utilities. NSTX will utilize the S-1 Spheromak vacuum vessel, poloidal field coils, and capacitor banks (for helicity injection). The poloidal Beta Experiment-Modified (PBX-M) power supplies will be shared to power the PF and TF coil systems. Existing RF hardware and infrastructure will be used for heating systems. TFTR data acquisition and diagnostics resources are planned to be used. In total, NSTX will utilize site credits with a value of {approximately}$50 M, reducing base construction cost of the device to $18.6 M. Twelve water-cooled copper demountable toroidal field (TF) coils provide the 5.4 kg (pulsed) and 3.5 kg (long pulse > 5 sec) toroidal field at the plasma center. Poloidal fields are generated by windings contained in the center column and four pairs water-cooled copper coils supported directly on the vacuum vessel. One of the most critical components of the device is the center stack, which consists of the inner legs of the TF coils overwrapped with ohmic heating and poloidal field windings. The ohmic heating coil windings are designed to optimize the V-s and together with the PF coils, produce a flux swing of 1 V-s.

  2. The JT-60 tokamak machine

    International Nuclear Information System (INIS)

    JT-60 is a large tokamak experimental device under construction at JAERI with main device parameters of R=3.0m, a=0.95m, Bsub(t)=45kG, and Isub(p)=2.7Ma. Its basic aim is to produce and confine hydrogen plasmas of temperatures in a multi-keV range and of confinement times comparable to a second, and to study its plasma-physics properties as well as engineering problems associated with them. The JT-60 tokamak machine is mainly composed of a vacuum vessel, toroidal field (TF) coils, poloidal field (PF) coils, and support structures. The vacuum vessel is a high toroidal chamber with an egg-shaped crossection, consisting of sectorial rigid rings and parallel bellows made from Inconel 625. It is baked out at a maximum temperature up to 5000C. Several kinds of first walls made from molybdenum are bolt-jointed to the vacuum vessel for its protection. The vacuum vessel is almost completely finished with design and is deeply into manufacturing. The TF system consists of 18 unit coils located around a torus axis at regular intervals. The unit coil composed of two pancakes are wedge-shaped at the section close to a torus axis and encased in a high-manganese non-magnetic steel case. Fabrication of the TF coils will be finished in May 1981. The PF coils are composed of ohmic heating coils, vertical field coils, horizontal field coils, and quadrupole field coils located inside the TF coil bore and outside the vacuum vessel, and magnetic limiter coils placed in the vacuum vessel. Its mechanical and thermal design is almost completed are composed of the upper and lower support structures, support comuns of the vacuum vessel, and central column made from high-manganese non-magnetic steel. The structural analysis was completed including a seismic analysis and the fabrication is now in progress. The first plasma is expected to be produced in October 1984. (orig.)

  3. Heavy Neutral Beam Probe for edge plasma analysis in Tokamaks

    International Nuclear Information System (INIS)

    The contents of this report present the progress achieved to date on the Heavy Neutral Beam Probe project. This effort is an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadien de Fusion Magnetique (CCFM). The overall objective of the effort is to develop and apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes (TdeV) facility in Montreal, Canada. To achieve this goal, a research and development project was established to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present the project is in the middle of its second budget period with the instrumentation on-site at TdeV. The first half of this budget period was used to complete total system tests at InterScience, Inc., dismantle and ship the hardware to TdeV, re-assemble and install the HNBP on the tokamak. Integration of the diagnostic into the TdeV facility has progressed to the point of first beam production and measurement on the plasma. At this time, the HNBP system is undergoing final de-bugging prior to re-start of machine operation in early Fall of this year

  4. Calculation of absorbed dose for skin contamination imparted by beta radiation through the VARSKIN modified code for 122 interesting isotopes for nuclear medicine, nuclear power plants and research

    International Nuclear Information System (INIS)

    In this work the implementation of a modification of the VARSKIN code for calculation of absorbed dose for contamination in skin imparted by external radiation fields generated by Beta emitting is presented. The modification consists on the inclusion of 47 isotopes of interest even Nuclear Plants for the dose evaluation in skin generated by 'hot particles'. The approach for to add these isotopes is the correlation parameter F and the average energy of the Beta particle, with relationship to those 75 isotopes of the original code. The methodology of the dose calculation of the VARSKIN code is based on the interpolation, (and integration of the interest geometries: punctual or plane sources), of the distribution functions scaled doses in water for beta and electrons punctual sources, tabulated by Berger. Finally a brief discussion of the results for their interpretation and use with purposes of radiological protection (dose insurance in relation to the considered biological effects) is presented

  5. Plasma Physics Regimes in Tokamaks with Li Walls

    International Nuclear Information System (INIS)

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors

  6. Small tokamaks for fusion technology testing

    International Nuclear Information System (INIS)

    Small steady-state tokamaks for testing divertors and fusion nuclear technologies are considered. Based on present physics and technology data and explanation to reduce R0/a, H-D-fueled tokamaks with R0 ∼ 0.6--0.75 m, R0/a ∼ 1.8--2.5, and Bt0 ∼ 1.4--2.2 T can be driven with Ptot ∼ 4.5 MW to maintain Ip ∼ 0.5 MA and produce the ITER-level plasma edge and divertor conditions. Given an adequate steady-state divertor solution and Q∼1 operation based on fusion through the suprathermal component, D-T-fueled tokamaks with R0 ∼ 0.8 m, R0/a ∼ 2, and Bt0 ∼ 4 T can be driven with Ptot ∼ 15 MW to maintain Ip ∼ 4.6 MA and produce an peak neutron wall load WL ∼ 1 MW/m2. Such devices appear possible if the plasma properties at the power R0/a remain tokamak-like and, for the D-T case, can unshielded center core is feasible. The use of a single conductor as the inboard leg of the toroidal field coils for this purpose is discussed. The physics issues and the design features are identified for such tokamaks with a testing duty for factor goal of 10--20%

  7. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

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

  8. Three novel tokamak plasma regimes in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Furth, H.P.

    1985-10-01

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

  9. Electron thermal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    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 (108 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)

  10. Simulation of burning tokamak plasmas

    International Nuclear Information System (INIS)

    To simulate dynamical behaviour of tokamak fusion reactors, a zero-dimensional time-dependent particle and power balance code has been developed. The zero-dimensional plasma model is based on particle and power balance equations that have been integrated over the plasma volume using prescribed profiles for plasma parameters. Therefore, the zero-dimensional model describes the global dynamics of a fusion reactor. The zero-dimensional model has been applied to study reactor start-up, and plasma responses to changes in the plasma confinement, fuelling rate, and impurity concentration, as well as to study burn control via fuelling modulation. Predictions from the zero-dimensional code have been compared with experimental data and with transport calculations of a higher dimensionality. In all cases, a good agreement was found. The advantage of the zero-dimensional code, as compared to higher-dimensional transport codes, is the possibility to quickly scan the interdependencies between reactor parameters. (88 refs., 58 figs., 6 tabs.)

  11. THOR tokamak magnetic field system

    International Nuclear Information System (INIS)

    The THOR Machine is an iron cored Tokamak having a major radius of 0.52 m and a minor radius of 0.17 m giving an aspect ratio of 3:1. It has a low ripple toroidal field of 1 T and an iron core giving 0.24 Vs. The maximum plasma current is expected to be in the region of 80x103 A. The maximum toroidal field ripple on axis is of the order of 0.01% and 2.5% at the plasma edge. The equilibrium of the plasma is achieved by means of a D.C. vertical field and a 1 cm thick copper shell. The D.C. field is cancelled during the rise time of the plasma current by means of pulsed reverse vertical field windings placed between the copper shell and the vacuum vessel. The design of this field system represents a compromise between obtaining adequate field penetration through the relatively thin vacuum vessel and maintaining the mechanical strength necessary to withstand the transient magnetic forces. Energy for the toroidal field system is supplied by a 15 kV 600 kJ capacitor bank and for the ohmic heating and reverse vertical fields by 5 kV 25 kJ and 50 kJ banks respectively. The problems encountered in the design, development and manufacture of these field systems are discussed. (author)

  12. Stability analysis of tokamak plasmas

    International Nuclear Information System (INIS)

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  13. Microtearing modes in tokamak discharges

    Science.gov (United States)

    Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.

    2016-06-01

    Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.

  14. Tokamak x ray diagnostic instrumentation

    International Nuclear Information System (INIS)

    Three classes of x-ray diagnostic instruments enable measurement of a variety of tokamak physics parameters from different features of the x-ray emission spectrum. (1) The soft x-ray (1 to 50 keV) pulse-height-analysis (PHA) diagnostic measures impurity concentrations from characteristic line intensities and the continuum enhancement, and measures the electron temperature from the continuum slope. (2) The Bragg x-ray crystal spectrometer (XCS) measures the ion temperature and neutral-beam-induced toroidal rotation velocity from the Doppler broadening and wavelength shift, respectively, of spectral lines of medium-Z impurity ions. Impurity charge state distributions, precise wavelengths, and inner-shell excitation and recombination rates can also be studied. X rays are diffracted and focused by a bent crystal onto a position-sensitive detector. The spectral resolving power E/ΔE is greater than 104 and time resolution is 10 ms. (3) The x-ray imaging system (XIS) measures the spatial structure of rapid fluctuations (0.1 to 100 kHZ) providing information on MHD phenomena, impurity transport rates, toroidal rotation velocity, plasma position, and the electron temperature profile. It uses an array of silicon surface-barrier diodes which view different chords of the plasma through a common slot aperture and operate in current (as opposed to counting) mode. The effectiveness of shields to protect detectors from fusion-neutron radiation effects has been studied both theoretically and experimentally

  15. Measurement 20-200 keV hard X-ray based on CdTe detector in EAST Tokamak

    International Nuclear Information System (INIS)

    Background: Accurate and quantitative measurement of plasma radiation is a key issue to Tokamak, toroidal magnetic confinement device. The radiations from Tokamak cover large energy range. Driven by the determination of the obtaining of hard X-ray spectra, a new system based on a high performance CdTe detector was built up in EAST Tokamak, the first non-circle cross-section in the world. Purpose: Introduces the device of hard X-ray diagnosis system in the EAST Tokamak on the Port A. The system can measure the plasma hard X-ray (20-200 keV) spectra under different discharge conditions, including Ohmical shot and Lower Hybrid Current Drive (LHCD) shot. The research of high speed electron which produced by LHCD is also the aim of the new system. Methods: A high performance CdTe detector was using in EAST Tokamak to measure the hard X-ray (20-200 keV) spectra. Results: The results show that the new system based on a high performance CdTe can meet the requirements for measuring the EAST Tokamak. Conclusions: A preliminary experimental result showed that the system can meet the requirements for measuring the X-ray bremsstrahlung of plasma in the energy range from 20 to 200 keV Calibration result and typical measurement result on EAST are present in this paper. (authors)

  16. The conceptual design of a robust, compact, modular tokamak reactor based on high-field superconductors

    Science.gov (United States)

    Whyte, D. G.; Bonoli, P.; Barnard, H.; Haakonsen, C.; Hartwig, Z.; Kasten, C.; Palmer, T.; Sung, C.; Sutherland, D.; Bromberg, L.; Mangiarotti, F.; Goh, J.; Sorbom, B.; Sierchio, J.; Ball, J.; Greenwald, M.; Olynyk, G.; Minervini, J.

    2012-10-01

    Two of the greatest challenges to tokamak reactors are 1) large single-unit cost of each reactor's construction and 2) their susceptibility to disruptions from operation at or above operational limits. We present an attractive tokamak reactor design that substantially lessens these issues by exploiting recent advancements in superconductor (SC) tapes allowing peak field on SC coil > 20 Tesla. A R˜3.3 m, B˜9.2 T, ˜ 500 MW fusion power tokamak provides high fusion gain while avoiding all disruptive operating boundaries (no-wall beta, kink, and density limits). Robust steady-state core scenarios are obtained by exploiting the synergy of high field, compact size and ideal efficiency current drive using high-field side launch of Lower Hybrid waves. The design features a completely modular replacement of internal solid components enabled by the demountability of the coils/tapes and the use of an immersion liquid blanket. This modularity opens up the possibility of using the device as a nuclear component test facility.

  17. Extremely shaped plasmas to improve the Tokamak concept

    International Nuclear Information System (INIS)

    experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is

  18. Simulation of runaway electrons in Tokamak disruptions

    International Nuclear Information System (INIS)

    Self-consistent modelling of the generation of runaway electrons and the evolution of the toroidal electric field during tokamak disruptions is presented. The process of runaway generation is analysed by combining a relativistic kinetic equation for the electrons with Maxwell's equations for the electric field. Such modelling allows for a quantitative assessment of the runaway generation during disruptions in present day tokamak experiments, and to extrapolate to future tokamaks like ITER. It is found that the current profile can change dramatically during a disruption, such that the post disruption current, carried mainly by the runaway electrons, is significantly more peaked than the current profile before the disruption. In fact, it is found that the central current density can increase in spite of a reduction in the total current. (authors)

  19. Activation analysis of the compact ignition tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak.

  20. Activation analysis of the compact ignition tokamak

    International Nuclear Information System (INIS)

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

  1. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

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

  2. Mass spectrometry instrumentation in TN (Novillo Tokamak)

    International Nuclear Information System (INIS)

    The mass spectrophotometry in the residual gases analysis in high vacuum systems, in particular in the Novillo Tokamak (TN), where pressures are required to be of the order 10-7 Torr, is carried out through an instrumental support with infrastructure configured in parallel to the experimental planning in this device. In the Novillo as well as other Tokamaks, it is necessary to condition the vacuum chamber for improving the main discharge parameters. At the present time, in this Tokamak the conditioning quality is presented determined by means of a mass spectrophotometer. A general instrumental description is presented associated with the Novillo conditioning, as well as the spectras obtained before and after operation. (Author)

  3. Research Advances on Acupuncture-moxibustion for Improving Pancreatic Beta-cell Function%针灸改善胰岛β细胞功能研究进展

    Institute of Scientific and Technical Information of China (English)

    庞晓英; 孙志

    2014-01-01

    2型糖尿病( T2DM)是临床常见多发病,严重危害人们的身体健康。胰岛β细胞功能障碍是T2 DM发生发展的中心环节。针灸治疗T2 DM有较好疗效,可以有效改善胰岛β细胞功能。笔者查阅了近15年针灸治疗糖尿病的相关文献资料,从实验研究和临床研究两方面对针灸改善胰岛β细胞功能的研究进行综述。%The type 2 diabetes mellitus ( T2DM) is a clinical common and frequently -occurring disease .It is seriously detrimental to people ’ s health.The dysfunction of pancreatic beta -cells is central to the occurrence and development of T2DM.Acupuncture -moxibustion treatment for T2DM obtains an curative effect.It can effectively protect the pancreatic beta -cells.We have consulted pertinent literatures of last 15 years, and will review researches on acupuncture -moxibustion for improving pancreatic beta -cell function from two sides of experimental and clinical researches .

  4. Ideal and resistive MHD stability of one-dimensional tokamak equilibria

    International Nuclear Information System (INIS)

    The MHD equilibrium and stability of a vertically elongated tokamak configuration are analyzed in the one-dimensional limit corresponding to infinite elongation. Stability against all ideal MHD modes can be obtained for beta-values arbitrarily close to unity. In the finite-resistivity stability analysis, axisymmetric (m = 0) tearing modes, centered on the null trace of the poloidal field, can be stabilized by a loosely fitting conducting shell. The presence of the toroidal field component, however, introduces the possibility of nonsymmetric tearing modes (m not equal to 0), centered away from the null trace. These modes can be stabilized only by a more tightly-fitting shell, plus reliance on finite-pressure effects on the small-major-radius side of the plasma profile. Under these conditions, stable configurations with peak beta values approaching unity are readily found

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

    International Nuclear Information System (INIS)

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

  6. Plasma equilibrium and field diffusion during current rise phase of STP-2 screw pinch tokamak

    International Nuclear Information System (INIS)

    Plasma equilibrium and field diffusion during the current rise phase of the discharge have been investigated in STP-2 screw pinch tokamak. The plasma with maximum poloidal beta value βsub(p) of 3.0 has been obtained by compression and joule heating. However the maximum βsub(p) value without strong wall contacts was about 1.3. We observed that force-free current is formed in the periphery of the plasma and the penetration rate of the poloidal magnetic field is much faster than the penetration rate calculated from the classical resistivity. To understand the high-beta plasma equilibrium and the mechanism of fast penetration rate observed in STP-2 plasmas we have performed a numerical simulation using a 2-D MHD pinch code ''TOPICS''. It is demonstrated that the fast penetration rate can be explained by introducing the influx of neutral particles and the ion acoustic type anomalous resistivity. (author)

  7. Current drive by spheromak injection into a tokamak

    OpenAIRE

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

    1990-01-01

    We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma (n¯3 increases by a factor of 6) then becomes hollow, suggestive of...

  8. Periodic disruptions in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Disruptive instabilities are common phenomena in toroidal devices, especially in tokamaks. Three types can be distinguished: internal, minor and major disruptions. Periodic minor disruptions in the MT-1 tokamak were measured systematically with values of the limiter safety factor between 4 and 10. The density limit as a function of plasma current and horizontal displacement was investigated. Precursor oscillations always appear before the instability with increasing amplitude but can be observed at the density limit with quasi-stationary amplitude. Phase correlation between precursor oscillations were measured with Mirnov coils and x-ray detectors, and they show good agreement with a simple magnetic island model. (R.P.) 11 refs.; 6 figs

  9. Can better modelling improve tokamak control?

    International Nuclear Information System (INIS)

    The control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed. (author) 5 figs., 9 refs

  10. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  11. A method for tokamak neutronics calculations

    International Nuclear Information System (INIS)

    This paper presents a new method for neutron transport calculation in tokamak fusion reactors. The computational procedure is based on the solution of the even-parity transport equation in a toroidal geometry. The angular neutron distribution is treated by even-parity spherical harmonic expansion, while the spatial dependence is approximated by using R-function finite elements that are defined for regions of arbitrary geometric shape. In order to test the method, calculation of a simplified tokamak model is carried out. The results are compared with the results from the literature and for the same order of accuracy a reduction of the number of spatial unknowns is shown. (author)

  12. Electronic system of TBR tokamak device

    International Nuclear Information System (INIS)

    The electronics developed as a part of the TBR project, which involves the construction of a small tokamak at the Physics Institute of the University of Sao Paulo, is described. On the basis of tokamak parameter values, the electronics for the toroidal field, ohmic/heating and vertical field systems is presented, including capacitors bank, switches, triggering circuits and power supplies. A controlled power oscilator used in discharge cleaning and pre-ionization is also described. The performance of the system as a function of the desired plasma parameters is discussed. (Author)

  13. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

  14. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    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

  15. Tokamak Spectroscopy for X-Ray Astronomy

    Science.gov (United States)

    Fournier, Kevin B.; Finkenthal, M.; Pacella, D.; May, M. J.; Soukhanovskii, V.; Mattioli, M.; Leigheb, M.; Rice, J. E.

    2000-01-01

    This paper presents the measured x-ray and Extreme Ultraviolet (XUV) spectra of three astrophysically abundant elements (Fe, Ca and Ne) from three different tokamak plasmas. In every case, each spectrum touches on an issue of atomic physics that is important for simulation codes to be used in the analysis of high spectral resolution data from current and future x-ray telescopes. The utility of the tokamak as a laboratory test bed for astrophysical data is demonstrated. Simple models generated with the HULLAC suite of codes demonstrate how the atomic physics issues studied can affect the interpretation of astrophysical data.

  16. Multichannel submillimeter interferometer for tokamak density measurements

    International Nuclear Information System (INIS)

    A two-channel, submillimeter (SMM) laser, electron-density interferometer has been operated successfully on the ISX tokamak. The interferometer is the first phase of a diagnostic system to measure the tokamak plasma current density using the Faraday rotation of the polarization vector of SMM laser beams. Deuterated formic acid lasers (lambda = 0.381 mm) have produced cw power of 10 mW. The interferometer has performed successfully for line-averaged electron densities as high as 8 x 1013 cm-3

  17. Tokamak power systems studies, FY 1985

    International Nuclear Information System (INIS)

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs

  18. 50 years forward: beta cells.

    Science.gov (United States)

    Halban, Philippe A

    2015-08-01

    Our understanding of beta cell development and function has increased substantially these past 50 years but much remains to be learned before this knowledge can be put to clinical use. A comprehensive business plan will be necessary to develop a detailed molecular and functional blueprint of the beta cell in health and disease based on an integrated approach involving all necessary research disciplines. This blueprint will provide a platform for the development of novel therapeutic strategies for the treatment of both major forms of diabetes, foremost among them beta cell replacement therapy. This is one of a series of commentaries under the banner '50 years forward', giving personal opinions on future perspectives in diabetes, to celebrate the 50th anniversary of Diabetologia (1965-2015). PMID:25957776

  19. Levered and unlevered Beta

    OpenAIRE

    Fernandez, Pablo

    2003-01-01

    We prove that in a world without leverage cost the relationship between the levered beta ( L) and the unlevered beta ( u) is the No-costs-of-leverage formula: L = u + ( u - d) D (1 - T) / E. We also analyze 6 alternative valuation theories proposed in the literature to estimate the relationship between the levered beta and the unlevered beta (Harris and Pringle (1985), Modigliani and Miller (1963), Damodaran (1994), Myers (1974), Miles and Ezzell (1980), and practitioners) and prove that all ...

  20. Dynamics and Feedback Control of Plasma Equilibrium Position in a Tokamak.

    Science.gov (United States)

    Burenko, Oleg

    A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems. The major parameters governing the plasma equilibrium position stability of a tokamak are shown to be (1) external magnetic field decay index, (2) transformer iron core effect, (3) plasma current, (4) radial rate-of-change inductance parameter, (5) vertical rate-of-change inductance parameter, and (6) vacuum vessel eddy-current time constant. An important and unique result is derived, showing that for a vacuum vessel eddy-current time constant exceeding a certain value the vertical plasma equilibrium position is stable, in spite of an intentional vertical instability design represented by a negative decay index. It is shown that a tokamak design having a theoretical set of positive decay index, negative radical rate-of-change inductance parameter, and positive vertical rate-of-change inductance parameter is expected to have a better plasma equilibrium position stability tolerance than a tokamak design having the same set with the signs reversed. The results of an actual hardware ISX-A tokamak plasma displacement feed-back control system design are presented. It is shown that a theoretical design computer

  1. Betting Against Beta

    DEFF Research Database (Denmark)

    Frazzini, Andrea; Heje Pedersen, Lasse

    .S. equities, 20 international equity markets, Treasury bonds, corporate bonds, and futures; (2) A betting-against-beta (BAB) factor, which is long leveraged low beta assets and short high-beta assets, produces significant positive risk-adjusted returns; (3) When funding constraints tighten, the return of the...

  2. Forward-Looking Betas

    DEFF Research Database (Denmark)

    Christoffersen, Peter; Jacobs, Kris; Vainberg, Gregory

    Few issues are more important for finance practice than the computation of market betas. Existing approaches compute market betas using historical data. While these approaches differ in terms of statistical sophistication and the modeling of the time-variation in the betas, they are all backward-...

  3. Scintillator based beta batteries

    Science.gov (United States)

    Rensing, Noa M.; Tiernan, Timothy C.; Shirwadkar, Urmila; O'Dougherty, Patrick; Freed, Sara; Hawrami, Rastgo; Squillante, Michael R.

    2013-05-01

    Some long-term, remote applications do not have access to conventional harvestable energy in the form of solar radiation (or other ambient light), wind, environmental vibration, or wave motion. Radiation Monitoring Devices, Inc. (RMD) is carrying out research to address the most challenging applications that need power for many months or years and which have undependable or no access to environmental energy. Radioisotopes are an attractive candidate for this energy source, as they can offer a very high energy density combined with a long lifetime. Both large scale nuclear power plants and radiothermal generators are based on converting nuclear energy to heat, but do not scale well to small sizes. Furthermore, thermo-mechanical power plants depend on moving parts, and RTG's suffer from low efficiency. To address the need for compact nuclear power devices, RMD is developing a novel beta battery, in which the beta emissions from a radioisotope are converted to visible light in a scintillator and then the visible light is converted to electrical power in a photodiode. By incorporating 90Sr into the scintillator SrI2 and coupling the material to a wavelength-matched solar cell, we will create a scalable, compact power source capable of supplying milliwatts to several watts of power over a period of up to 30 years. We will present the latest results of radiation damage studies and materials processing development efforts, and discuss how these factors interact to set the operating life and energy density of the device.

  4. Recent results from DIII-D and their implications for next generation tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Luxon, J.L.; Bramson, G.; Burrell, K.H.; Brooks, N.H.; Callis, R.W.; Carlstrom, T.N.; Chu, M.S.; Colleraine, A.P.; DeBoo, J.C.; Ferron, J.R.; Freeman, R.; Gohil, P.; Greenfield, C.M.; Groebner, R.J.; Hong, R.; Howl, W.; Hsieh, C.L.; Jackson, G.L.; Jensen, T.; Kellman, A.G.; Kim, J.; Lao, L.L.; La Haye, R.; Leikind, B.; Lippmann, S.I.; Lohr, J.; Luce, T.C.; Mahdavi, M.A.; Mayberry, M.; Moeller, C.P.; Osborne, T.H.; Overskei, D.O

    1990-07-01

    Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements of tokamak physics and the application of this understanding to the design of next generation devices including ITER and CIT. The limitations of magnetohydrodynamics stability on the values of plasma beta (the ratio of kinetic pressure to the containing pressure of the magnetic field) that can be attained has been experimentally demonstrated and found to be described by existing theory. Values of beta (10.7%) well in excess of those required for proposed devices (ITER and CIT) have been demonstrated. Regimes of confinement (H-mode) have been established that scale favorably to proposed next generation devices, and experiments demonstrating the dependence of the energy confinement on plasma size have been completed. Understanding of confinement is rapidly developing especially in the areas of bulk transport and the role of turbulence in the plasma edge. Key experimental results in areas of plasma transport and edge plasma phenomena are found to be in agreement with theories based on short wavelength turbulence. Control of the divertor heat loads and impurity influx has been demonstrated, and new progress has been made in the understanding of plasma edge phenomena. Experiments with ion Bernstein wave heating have not found regimes in which these waves can produce effective central ion heating. Electron cyclotron current drive experiments have demonstrated 70 kA of driven current in 400 kA discharges.

  5. Deuterium-Tritium Simulations of the Enhanced Reversed Shear Mode in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D.R.; Manickam, J.; Scott, S.D.; Zarnstorff

    1997-04-01

    The potential performance, in deuterium-tritium plasmas, of a new enhanced con nement regime with reversed magnetic shear (ERS mode) is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle dif- fusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F. M. Levinton, et al., Phys. Rev. Letters, 75, 4417, (1995)]. The plasma performance depends strongly on Zeff and neutral beam penetration to the core. The steady state projections typically have a central electron density of {approx}2:5x10 20 m{sup -3} and nearly equal central electron and ion temperatures of {approx}10 keV. In time dependent simulations the peak fusion power, {approx} 25 MW, is twice the steady state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of {approx} 15 keV. The simulated pressure profiles can be stable to ideal MHD instabilities with toroidal mode number n = 1, 2, 3, 4 and {infinity} for {beta}{sub norm} up to 2.5; the simulations have {beta}{sub norm} {le} 2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions imilar to those proposed for advanced tokamak reactors.

  6. Studies on equilibrium analysis and cylindrical model of low aspect ratio tokamak

    International Nuclear Information System (INIS)

    The low-aspect-ratio tokamak is expected to realize a compact and economical fusion reactor. This paper presents a equilibrium study of low-aspect-ratio tokamaks for the basis of both of Magnetohydrodynamic (MHD) analysis and cylindrical approximation model. Since its external toroidal field coil does not have enough surrounding space for the installation of neutron shielding material, it should be made of a normal conductor, whose current density and the resistive power loss should be estimated. On the basis of the cylindrical approximation model, the relationships are derived between the resistive power loss and various plasma parameters as aspect ratio, elongation, safety factor at the edge, and total beta value. The resistive power loss can be decreased by decreasing aspect ratio, however, neutron wall loading and the total beta increase. It is found that they can be in the reasonable quantitative range by the appropriate selections of the plasma parameters. The evaluation is made of the applicability of the model by comparing the results with those of the numerical MHD analysis. (author)

  7. Alpha particle effects on global MHD modes, and alpha particle transport in ignited tokamaks

    International Nuclear Information System (INIS)

    The high frequency, low mode number toroidicity-induced Alfven eigenmodes (TAE) are shown to be driven unstable primarily by the circulating α-particles through wave-particle resonances. To destabilize the TAE modes, the inverse Landau damping associated with the α-particle pressure gradient free energy must overcome the velocity space Landau damping due to both the α-particles and the core electrons and ions, as well as Alfven continuum damping. Stability criteria are presented for TFTR, CIT, and ITER tokamaks in terms of the α-particle beta βα, the α-particle pressure gradient parameter (ω*/ωA), where ω* is the α-particle diamagnetic drift frequency, and the α-particle velocity (vα/vA) parameter. Typically the volume averaged α-particle beta threshold is on the order of 10-4. Rough estimates of the TAE mode saturation level give δBr/B ∼ 10-3 for typical D-T tokamak operations. Significant α-particle losses are found when the amplitude of the global MHD modes is large, on the order of (δBr/B) ≥ 10-4. For (δBr/B) = 5 x 10-4, the α-particle loss time is appreciably shorter than the α-particle slowing-down time. 13 refs., 1 fig

  8. On the fusion triple product and fusion power gain of tokamak pilot plants and reactors

    Science.gov (United States)

    Costley, A. E.

    2016-06-01

    The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτ E, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τ E are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτ E becomes almost independent of size; rather it depends mainly on the fusion power, P fus. In consequence, the fusion power gain, Q fus, a parameter closely linked to nTτ E is also independent of size. Hence, P fus and Q fus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτ E is inversely dependent on the normalised beta, β N; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

  9. A guide to 17beta-hydroxysteroid dehydrogenases.

    Science.gov (United States)

    Adamski, J; Jakob, F J

    2001-01-22

    17beta-Hydroxysteroid dehydrogenases (17beta-HSD) are pivotal in controlling the biological potency of steroid hormones by catalyzing oxidation or reduction at position 17. Several 17beta-HSDs may as well metabolize further substrates including alcohols, bile acids, fatty acids and retinols. This review summarizes recent progress in the field of 17beta-HSD research provides an update of nomenclature. PMID:11165003

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

    International Nuclear Information System (INIS)

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

  11. A Web-Based System for Remote Data Browsing in HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    Cheng Ting; Luo Jiarong; Meng Yuedong; Wang Huazhong

    2005-01-01

    HT-7 is the first superconducting tokamak device for fusion research in China. Many experiments have been performed on the HT-7 tokamak since 1994 with numerous satisfactory results achieved in the fusion research field. As more and better communication is required with other fusion research laboratories, remote access to experimental data is becoming increasingly important in order to raise the degree of openness of experiments and to expand research results.The web-based remote data browsing system enables authorized users in geographically different locations to view and search for experimental data without having to install any utility software at their terminals. The three-tier software architecture and thin client technology are used to operate the system effectively. This paper describes the structure of the system and the realization of its functions, focusing on three main points: the communication between the participating tiers, the data structure of the system and the visualization of the raw data on web pages.

  12. Annual report of the Fusion Research and Development Center for the period of April 1, 1980 to March 31, 1981

    International Nuclear Information System (INIS)

    Research and development activities of the Fusion Research and Development Center (Division of Thermonuclear Fusion Research and Division of Large Tokamak Development) from April 1980 to 1981 are described. In plasma physics research, 1.5 MW NBI heating experiments were successfully made on JFT-2 to yield an average beta value of 2.5% without any deleterious effect on plasma confinement. Joint JAERI-US/DOE ECRH experiments revealed detailed physics of plasma heating. Installation of a 1 MW ICRF system was completed. In the Doublet-III experiment, a JAERI-US/DOE cooperation program, extensive studies were made on Joule heated dee-shaped Plasmas. In theory and computation emphasis was placed on beta optimization of tokamaks. Construction of JT-60 was continued as planned. Manufacturing of the major components and facilities was advanced well, e.g. 14 out of the 19 toroidal field coils were completed. Construction of the buildings was continued at the Naka site. In plasma heating technology, construction of the JT-60 prototype NBI unit was in progress, and development works on ion sources and beam line components as well. Trial fabrication of high power klystrons for JT-60 RF heating was started. In superconducting magnet technology, cool-down tests of cluster coils were successfully made. Manufacturing of the Japanese coil for the Large Coil Task under the auspiece of IEA, and of a Nb3Sn test module coil was continued. A test facility for the LCT coil was completed. Basic studies on key processes of tritium technology were continued using hydrogen and deuterium. Design of the Tritium Process Laboratory was continued. Development of first wall materials for JT-60 was advanced. Extensive tests were made on a number of low-Z coatings. Design studies of INTOR, a cooperative work in IAEA, were continued. In addition, design of the Fusion Experimental Reactor was started on a conventional type tokamak reactor and swimming pool type one. (author)

  13. Atomic data for integrated tokamak modelling

    International Nuclear Information System (INIS)

    . Moreover we present elastic cross sections of fusion related materials. We present total and angular differential elastic cross sections of hydrogen atoms for a wide range of incident electron energy. One of the convenient ways of representation of these data is analytical fit functions, which can be easily applied in various fields of sciences. The aim of work is to develop a universal functional formula of elastic cross sections for the case of hydrogen target. We consider the angular differential electron elastic cross sections for a wide range of incident electron energy and in the entire angular range. The differential cross-sections were calculated using the partial expansion method. The fitted curves are in excellent agreement with the calculated ones within less than 1% accuracy in the energy range between 1 eV and 100 keV. Our analytical formula may show the main virtues in various Monte Carlo simulations reducing drastically the computation time, when it requires to calculate the elastic cross sections many times. The applied fitting technique can be used for other data. Acknowledgement: This work, supported by the European Communities under the contract of Association between EURATOMHAS, was carried out within the framework of the Task Force on Integrated Tokamak Modeling of the European Fusion Development Agreement. The work was also supported by the Hungarian Scientific Research Fund OTKA No. NN103279. (author)

  14. Microinstabilities in weak density gradient tokamak systems

    International Nuclear Information System (INIS)

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient

  15. High βp bootstrap tokamak reactor

    International Nuclear Information System (INIS)

    Basic characteristics of a steady state tokamak fusion reactor is presented. The minimum required energy multiplication factor Q is found to be 20 to 30 for the feasibility of the fusion reactor. Such a high Q steady state tokamak operation is possible, within our present knowledge of the operational constraints and the current drive physics, when a large fraction of the plasma current is carried by the bootstrap current. Operation at high βp (≥2.0) and high qψ (=4-5) with relatively small εβp (3) and fusion output power (2.5 GW) and is consistent with the present knowledges of the plasma physics of the tokamak, namely the Troyon limit, the energy confinement scalings, the bootstrap current, the current drive efficiency (NB current drive with the total power of 70 MW and the beam energy of 1 MeV) with a favorable aspect on the formation of the cold and dense diverter plasma-condition. From the economical aspect of the tokamak fusion reactor, a more compact reactor is favorable. The use of the high field magnet with Bmax = 16T (for example Ti-doped Nb3Sn conductor) enables to reduce the total machine size to 50% of the above-described conventional design, namely Rp = 7m, Vp = 760m-3, PF = 2.8 GW. (author)

  16. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  17. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  18. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  19. Toroidal Alfven wave stability in ignited tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, C.Z.; Fu, G.Y.; Van Dam, J.W.

    1989-01-01

    The effects of fusion-product alpha particles on the stability of global-type shear Alfven waves in an ignited tokamak plasma are investigated in toroidal geometry. Finite toroidicity can lead to stabilization of the global Alfven eigenmodes, but it induces a new global shear Alfven eigenmodes, which is strongly destabilized via transit resonance with alpha particles. 8 refs., 2 figs.

  20. Analysis of sawtooth relaxation oscillations in tokamaks

    International Nuclear Information System (INIS)

    Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated

  1. Tokamak power systems studies: A second stability power reactor

    International Nuclear Information System (INIS)

    A number of innovative physics and engineering features have been studied which promise to greatly improve the reactor prospects of tokamaks relative to STARFIRE. A reference design point has been developed with the following features: large aspect ratio (A = 6); high beta (β ≅ 0.20), with only mild shaping and no indentation, which brings the maximum toroidal field down to 7 T; low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and EF coil system; and steady-state operation with combined fast wave and lower hybrid wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with control of the current achieved by the appropriate choice of wave frequencies and spectra. By selecting an axial safety factor q(o) = 2.0, MHD stability has been found above β ≅ 0.20. Additional features include: impurity control with self-pumped limiters which bury helium on continuously deposited metal surfaces; liquid Li-cooled blanket which provides good performance with low pressure operation; vanadium alloy blanket structure for higher thermal efficiency (eta = 0.42), longer lifetime and reduced activation; and reduced reactor mass (higher power density) due to smaller TF coil, less shielding, fewer blanket penetrations, and higher wall loading. At low neutron wall loads this device represents a minimum capital cost unit. However, economies of scale are strong, and eventually higher wall loads (W ≅ 8 MW/m2, P/sub net/ = 1400 MW) may prove most attractive. Preliminary investigations show inherently safe operation is likely at W ≥ 5 MW/m2. 15 refs., 3 figs., 1 tab

  2. Slow bank system of SINP-Tokamak: A short report

    International Nuclear Information System (INIS)

    SINP Tokamak was made operational in July, 1987. The power supply system of the tokamak at that time was designed for a plasma duration of around 2 ms for a peak plasma current of 75 kA. Efforts were directed to increase this duration to 20 ms with the help of a slow bank system designed to work in conjunction with the original fast bank system. The design aspects of the system were completed and the system has been partially executed. Subsequent to this partial implementation, efforts were directed to incorporate the necessary control system and interface facilities between the existing fast bank and the developed slow bank systems. The significant features of the control circuits are that they work according to a well thought out sequences of logic and are designed to guard against possible failures in the existing or the developed power supplies. Efforts have been put to make the operation of the system as much user-friendly as could be worked out within certain practical constraints. The control circuit and interface facilities have been put to extensive tests and are found to work satisfactorily. The entire power supply system is now in active use for different research programmes in the group. (author)

  3. Observation of MHD phenomenon for SST-1 superconducting tokamak

    International Nuclear Information System (INIS)

    Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (major radius = 1.1 m, minor radius = 0.2 m) and is operational at the Institute for Plasma Research (IPR), India. In the last few experimental campaigns SST-1 has successfully achieved plasma current in order of 60-70 kA and plasma duration in excess of ∼ 500ms at a central magnetic field of 1.5T. An attempt has made to study the behavior of the magneto-hydrodynamic (MHD) activity during different phases of plasma pulse which leads to major/minor disruptions, its present modes (poloidal/toroidal mode number i.e. m=2, n=1) impact on plasma confinement and signature of lock mode and its frequency in the SST-1 plasma using experimental data from Mirnov signals. Observed MHD phenomenon has also been correlated with other diagnostics (i.e. ECE, Density, X-Ray etc.) and heating system (ECRH) for the recent campaigns of SST-1. (author)

  4. Ohmic discharges with improved confinement in Tokamak Aditya

    International Nuclear Information System (INIS)

    ADITYA (R0 = 75 cm, a = 25 cm), an ohmically heated circular limiter tokamak is regularly being operated to carry out several experiments related to controlled thermonuclear fusion research. In recent experimental schedule, special efforts are made to enhance the plasma parameters to achieve Ohmic discharges with improved confinement. Repeatable plasma discharges of maximum plasma current of ∼ 160 kA and discharge duration beyond ∼ 250 ms with plasma current flattop duration of ∼ 140 ms has been obtained for the first time in the first Indian tokamak ADITYA. The discharge reproducibility has been improved with Lithium wall conditioning and much-improved plasma discharges are obtained by precisely controlling the plasma position. Improved discharges are attempted over a wider parameter range to carry out various confinement scaling experiments. In these discharges, chord-averaged electron density 1.0 - 4.0 X 1019m-3 using multiple hydrogen gas puffs, plasma temperature of the order of ∼ 400 - 700 eV has been achieved. The measured confinement time matches quite well with ALCATOR scaling for most of the discharges. It is also observed that in new discharges, the confinement time crosses the L-mode scaling. Detailed analysis of these discharges along with the possible reasons for obtaining higher confinement times will be addressed in this paper. (author)

  5. Measurement of Sheared Flows in the Edge Plasma of the CASTOR Tokamak

    Czech Academy of Sciences Publication Activity Database

    Brotánková, Jana; Stöckel, Jan; Horáček, Jan; Seidl, Jakub; Ďuran, Ivan; Hron, Martin; Van Oost, G.

    2009-01-01

    Roč. 35, č. 11 (2009), s. 980-986. ISSN 1063-780X. [IAEA Technical Meeting on Research Using Small Fusion Devices/18th./. Alushta (Krym), 25.09.2008-27.09.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * probe diagnostics * shear ed flows * edge plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.584, year: 2009 http://www.springerlink.com/content/u571504gmq118314/

  6. A conceptual design of superconducting spherical tokamak reactor

    International Nuclear Information System (INIS)

    This paper presents a fusion reactor concept named 'JUST (Japanese Universities' Super Tokamak reactor)'. From the plasma confinement system to the power generation system is evaluated in this work. JUST design has features as follows: the superconducting magnet, the steady state operation with high bootstrap current fraction, the easy replacement of neutron damaged first wall, the high heat flux in the divertor, and the low cost (or high β). By winding the OH solenoid over the center stack of toroidal field coil, we have the low aspect ratio and the 80cm thick neutron shield to protect the superconducting center stack. JUST is designed by using the 0-D transport code under the assumption that the energy confinement time is 1.8 times of the IPB98(y,2) scaling. Main parameters are as follows: the major radius of 4.5m, the aspect ratio of 1.8, the elongation ratio of 2.5, the toroidal field of 2.36T, the plasma current of 18MA, the toroidal beta of 22%, the central electron and ion temperature of 15keV and the fusion thermal power of 2.4GW. By using the mercury heat exchanger and the steam turbine, the heat efficiency is 33% and the electric power is 0.74GW. (author)

  7. Comprehensive gyrokinetic simulation of tokamak turbulence at finite relative gyroradius

    International Nuclear Information System (INIS)

    A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate turbulent transport in actual experimental profiles and allow direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite beta, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius (ρ*) so as to treat the profile shear stabilization effects which break gyro Bohm scaling. The code operates in a cyclic flux tube limit which allows only gyro Bohm scaling and a noncylic radial annulus with physical profile variation. The later requires an adaptive source to maintain equilibrium profiles. Simple ITG simulations demonstrate the broken gyro Bohm scaling paradigm of Garbet and Waltz [Phys. Plasmas 3, 1898 (1996)]. Since broken gyro Bohm scaling depends on the actual rotational velocity shear rates competing with mode growth rates, direct comprehensive simulations of the DIII-D ρ*-scaled L-mode experiments are presented as a quantitative test of gyrokinetics and the paradigm. (author)

  8. International tokamak reactor. Phase Two A, Part 3. Vol.2

    International Nuclear Information System (INIS)

    Volume 2 of the report of Phase 2A, Part III of the International Tokamak Reactor Workshop focuses on a critical analysis of five INTOR-like designs: INTOR, NET, FER, TIBER, and OTR. Topics covered are a report of IAEA Specialists' Meeting on INTOR-like designs, a comparison of these designs, and a systems analysis of the designs. In this volume, programmatic and technical objectives are discussed and compared. The resulting physics design features are based on Q (power out/power in), burn duration, impurity control, plasma configuration, startup, operational scenario, duty factor, plasma heating, and frequency of disruptions. Design constraints (confinement scaling, plasma beta, density limit, geometry, plasma parameters, operating mode, disruption, magnetic field, heating, and startup) are compared for the various machines. Systems analyses are used to determine sensitivities to the various physical design features, and those parameters with the most sensitivity are determined. Relative costs of the various designs are compared, ranging from TIBER at 35 per cent less than INTOR, to OTR at 25 per cent more than INTOR. Refs, figs and tabs

  9. High βp plasma formation using off-axis ECCD in Ohmic heated plasma in the spherical tokamak QUEST

    Directory of Open Access Journals (Sweden)

    Mishra Kishore

    2015-01-01

    Full Text Available High poloidal beta (ɛβp ~ 1 operation in steady state condition in tokamaks is of great interest and has previously been demonstrated using NBI, LHCD and low current (Ip plasma for a short time (<0.5 s. A very few experiments however, have been performed towards the investigation of highest obtainable βp in tokamak plasma. In this work we report the first result of high βp production and its sustainment though an off axis ECCD at two different frequencies (fundamental and second harmonic in Ohmic (OH target plasma. With application of ECCD, plasma βp increased to encounter an equilibrium limit and the standard limiter configuration is transformed to an Inboard Poloidal field Null (IPN configuration. Both off-axis and on-axis ECCD is studied and found to have some distinctive features, which are discussed in this paper.

  10. Systematic Risk on Istanbul Stock Exchange: Traditional Beta Coefficient Versus Downside Beta Coefficient

    Directory of Open Access Journals (Sweden)

    Gülfen TUNA

    2013-03-01

    Full Text Available The aim of this study is to test the validity of Downside Capital Asset Pricing Model (D-CAPM on the ISE. At the same time, the explanatory power of CAPM's traditional beta and D-CAPM's downside beta on the changes in the average return values are examined comparatively. In this context, the monthly data for seventy three stocks that are continuously traded on the ISE for the period 1991-2009 is used. Regression analysis is applied in this study. The research results have shown that D-CAPM is valid on the ISE. In addition, it is obtained that the power of downside beta coefficient is higher than traditional beta coefficient on explaining the return changes. Therefore, it can be said that the downside beta is superior to traditional beta in the ISE for chosen period.

  11. Equilibrium system analysis in a tokamak ignition experiment

    International Nuclear Information System (INIS)

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies

  12. Equilibrium system analysis in a tokamak ignition experiment

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  13. Equilibrium system analysis in a tokamak ignition experiment. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

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

    International Nuclear Information System (INIS)

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

  15. Maryland controlled fusion research program

    International Nuclear Information System (INIS)

    In this paper, we summarize the technical progress in four major areas of tokamak research: (a) L/H transition and edge turbulence and transport; (b) active control of microturbulence and transport; (c) major disruptions; and (d) the sawtooth crash

  16. Characterization of the Tokamak Novillo in cleaning regime; Caracterizacion del Tokamak Novillo en regimen de limpieza

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-02-15

    In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip{sub t} like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I{sub (p)}t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)

  17. First experiments on the TO-2 tokamak with a divertor

    International Nuclear Information System (INIS)

    Long stable discharges have been obtained in a recetrack tokamak with toroidal divertors in low plasma density regime. Divertors sharply limit plasma filament cross section, plasma density decreasing by an order at 1 cm length near the separatrix. 8 mm thick well formed flux of plasma appears at the divertor plate. Divertor power efficiency at different modes of operation is 50- 70 %. As compared to the TO-1 nondivertor tokamak some plasma filament hot zone expansion is recorded in the TO-2 tokamak

  18. Banana orbits in elliptic tokamaks with hole currents

    Science.gov (United States)

    Martin, P.; Castro, E.; Puerta, J.

    2015-03-01

    Ware Pinch is a consequence of breaking of up-down symmetry due to the inductive electric field. This symmetry breaking happens, though up-down symmetry for magnetic surface is assumed. In previous work Ware Pinch and banana orbits were studied for tokamak magnetic surface with ellipticity and triangularity, but up-down symmetry. Hole currents appear in large tokamaks and their influence in Ware Pinch and banana orbits are now considered here for tokamaks magnetic surfaces with ellipticity and triangularity.

  19. Steady State Advanced Tokamak (SSAT): The mission and the machine

    International Nuclear Information System (INIS)

    Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

  20. Systems studies of high-field tokamak ignition experiments

    International Nuclear Information System (INIS)

    A study of the interaction between the physics of ignition and the engineering constraints in the design of compact, high-field tokamak ignition demonstration devices is presented. The studies investigate the effects the various electron and ion thermal diffusivities, which result from the many tokamak scaling laws, have on the design parameters of an ignition device and show the feasibility of building and igniting a compact tokamak (R<1m). The relevant machine technology is discussed