Tokamak configuration analysis with the method of toroidal multipoles

International Nuclear Information System (INIS)

Micozzi, P.; Alladio, F.; Crisanti, F.; Marinucci, M.; Tanga, A.

1989-01-01

In the study of tokamak machines able to sustain plasmas of thermonuclear interest (JIT, IGNITOR, NET, CIT, ET), there is a strong quest for engineering optimization of the circuital components close to the plasma. We have developed a semianalytical axisymmetric MHD equilibrium code based on the technique of the poloidal ψ flux function expansion in toroidal harmonic series. This code is able to optimize the necessary currents in the poloidal circuits in order to sustain a plasma of fixed shape (also x-point configuration), toroidal current and poloidal β. (author) 4 refs., 4 figs

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

International Nuclear Information System (INIS)

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

1991-07-01

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

Resistive demountable toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

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

1981-07-01

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

Fuelling effect of tangential compact toroid injection in STOR-M Tokamak

Energy Technology Data Exchange (ETDEWEB)

Onchi, T.; Liu, Y., E-mail: tao668@mail.usask.ca [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Dreval, M. [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Inst. of Plasma Physics NSC KIPT, Kharkov (Ukraine); McColl, D. [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Asai, T. [Inst. of Plasma Physics NSC KIPT, Kharkov (Ukraine); Wolfe, S. [Nihon Univ., Dept. of Physics, Tokyo (Japan); Xiao, C.; Hirose, A. [Univ. of Saskatchewan, Saskatoon, Saskatchewan (Canada)

2012-07-01

Compact torus injection (CTI) is the only known candidate for directly fuelling the core of a tokamak fusion reactor. Compact torus (CT) injection into the STOR-M tokamak has induced improved confinement accompanied by an increase in the electron density, reduction in Hα emission, and suppression of the saw-tooth oscillations. The measured change in the toroidal flow velocity following tangential CTI has demonstrated momentum injection into the STOR-M plasma. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Neoclassical poloidal and toroidal rotation in tokamaks

International Nuclear Information System (INIS)

Kim, Y.B.; Diamond, P.H.; Groebner, R.J.

1991-01-01

Explicit expressions for the neoclassical poloidal and toroidal rotation speeds of primary ion and impurity species are derived via the Hirshman and Sigmar moment approach. The rotation speeds of the primary ion can be significantly different from those of impurities in various interesting cases. The rapid increase of impurity poloidal rotation in the edge region of H-mode discharges in tokamaks can be explained by a rapid steepening of the primary ion pressure gradient. Depending on ion collisionality, the poloidal rotation speed of the primary ions at the edge can be quite small and the flow direction may be opposite to that of the impurities. This may cast considerable doubts on current L to H bifurcation models based on primary ion poloidal rotation only. Also, the difference between the toroidal rotation velocities of primary ions and impurities is not negligible in various cases. In Ohmic plasmas, the parallel electric field induces a large impurity toroidal rotation close to the magnetic axis, which seems to agree with experimental observations. In the ion banana and plateau regime, there can be non-negligible disparities between primary ion and impurity toroidal rotation velocities due to the ion density and temperature gradients. Detailed analytic expressions for the primary ion and impurity rotation speeds are presented, and the methodology for generalization to the case of several impurity species is also presented for future numerical evaluation

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Science.gov (United States)

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

2018-02-01

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

Theory for neoclassical toroidal plasma viscosity in tokamaks

International Nuclear Information System (INIS)

Shaing, K C; Chu, M S; Hsu, C T; Sabbagh, S A; Seol, Jae Chun; Sun, Y

2012-01-01

Error fields and magnetohydrodynamic modes break toroidal symmetry in tokamaks. The broken symmetry enhances the toroidal plasma viscosity, which results in a steady-state toroidal plasma flow. A theory for neoclassical toroidal plasma viscosity in the low-collisionality regimes is developed. It extends stellarator transport theory to include multiple modes and to allow for |m − nq| ∼ 1. Here, m is the poloidal mode number, n is the toroidal mode number and q is the safety factor. The bounce averaged drift kinetic equation is solved in several asymptotic limits to obtain transport fluxes. These fluxes depend non-linearly on the radial electric field except for those in the 1/ν regime. Here, ν is the collision frequency. The theory is refined to include the effects of the superbanana plateau resonance at the phase space boundary and the finite ∇B drift on the collisional boundary layer fluxes. Analytical expressions that connect all asymptotic limits are constructed and are in good agreement with the numerical results. The flux–force relations that relate transport fluxes to forces are used to illustrate the roles of transport fluxes in the momentum equation. It is shown that the ambipolar state is reached when the momentum equation is relaxed. It is also shown that the origin of the momentum for plasma flow generated without momentum sources is the local unbalance of particles' momenta and is diamagnetic in nature regardless of the details of the theory. (paper)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

Schmalz, R.

1982-09-01

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

Collisional boundary layer analysis for neoclassical toroidal plasma viscosity in tokamaks

Czech Academy of Sciences Publication Activity Database

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

2008-01-01

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

Toroidal inhomogeneity of the vertical field in a tokamak apparatus

International Nuclear Information System (INIS)

Sometani, Taro; Takashima, Hidekazu

1977-01-01

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

Passing particle toroidal precession induced by electric field in a tokamak

International Nuclear Information System (INIS)

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

2013-01-01

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

Passing particle toroidal precession induced by electric field in a tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2013-12-15

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

Effects of orbit squeezing on neoclassical toroidal plasma viscosity in tokamaks

Czech Academy of Sciences Publication Activity Database

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

2008-01-01

Roč. 15, č. 8 (2008), 082505-1-082505-8 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma instability * plasma magnetohydrodynamics * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2965146

Toroidal microinstability studies of high temperature tokamaks

International Nuclear Information System (INIS)

Rewoldt, G.; Tang, W.M.

1989-07-01

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

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

International Nuclear Information System (INIS)

Ida, Katsumi

2001-01-01

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

Analysis of toroidal vacuum vessels for use in demonstration sized tokamak reactors

International Nuclear Information System (INIS)

Culbert, M.E.

1978-07-01

The vacuum vessel component of the tokamak fusion reactor is the subject of this study. The main objective of this paper was to provide guidance for the structural design of a thin wall externally pressurized toroidal vacuum vessel. The analyses are based on the available state-of-the-art analytical methods. The shortcomings of these analytical methods necessitated approximations and assumptions to be made throughout the study. A principal result of the study has been the identification of a viable vacuum vessel design for the Demonstration Tokamak Hybrid Reactor (DTHR) and The Next Step (TNS) Reactor

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

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

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

Science.gov (United States)

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

2012-10-01

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

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

International Nuclear Information System (INIS)

Mazzucato, E.

2000-01-01

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

Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta

International Nuclear Information System (INIS)

Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.

2017-01-01

Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t ), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.

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

International Nuclear Information System (INIS)

Mao Jianshan; Zhao Junyu; Shen Biao; Luo Jiarong

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Poloidal and toroidal heat flux distribution in the CCT tokamak

International Nuclear Information System (INIS)

Brown, M.L.; Dhir, V.K.; Taylor, R.J.

1990-01-01

Plasma heat flux to the Faraday shield panels of the UCLA Continuous Current Tokamak (CCT) has been measured calorimetrically in order to identify the dominant parameters affecting the spatial distribution of heat deposition. Three heating methods were investigated: audio frequency discharge cleaning, RF heating, and AC ohmic. Significant poloidal asymmetry is present in the heat flux distribution. On the average, the outer panels received 25-30% greater heat flux than the inner ones, with the ratio of maximum to minimum values attaining a difference of more than a factor of 2. As a diagnostic experiment the current to a selected toroidal field coil was reduced in order to locally deflect the toroidal field lines outward in a ripple-like fashion. Greatly enhanced heat deposition (up to a factor of 4) was observed at this location on the outside Faraday panels. The enhancement was greatest for conditions of low toroidal field and low neutral pressure, leading to low plasma densities, for which Coulomb collisions are the smallest. An exponential model based on a heat flux e-folding length describes the experimentally found localization of thermal energy quite adequately. (orig.)

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

International Nuclear Information System (INIS)

Tani, Keiji; Kobayashi, Tomofumi; Tamura, Sanae

1975-02-01

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

Evidence of Inward Toroidal Momentum Convection in the JET Tokamak

DEFF Research Database (Denmark)

Tala, T.; Zastrow, K.-D.; Ferreira, J.

2009-01-01

Experiments have been carried out on the Joint European Torus tokamak to determine the diffusive and convective momentum transport. Torque, injected by neutral beams, was modulated to create a periodic perturbation in the toroidal rotation velocity. Novel transport analysis shows the magnitude...... and profile shape of the momentum diffusivity are similar to those of the ion heat diffusivity. A significant inward momentum pinch, up to 20 m/s, has been found. Both results are consistent with gyrokinetic simulations. This evidence is complemented in plasmas with internal transport barriers....

Locked magnetic island chains in toroidally flow damped tokamak plasmas

International Nuclear Information System (INIS)

Fitzpatrick, R; Waelbroeck, F L

2010-01-01

The physics of a locked magnetic island chain maintained in the pedestal of an H-mode tokamak plasma by a static, externally generated, multi-harmonic, helical magnetic perturbation is investigated. The non-resonant harmonics of the external perturbation are assumed to give rise to significant toroidal flow damping in the pedestal, in addition to the naturally occurring poloidal flow damping. Furthermore, the flow damping is assumed to be sufficiently strong to relax the pedestal ion toroidal and poloidal fluid velocities to fixed values determined by neoclassical theory. The resulting neoclassical ion flow causes a helical phase-shift to develop between the locked island chain and the resonant harmonic of the external perturbation. Furthermore, when this phase-shift exceeds a critical value, the chain unlocks from the resonant harmonic and starts to rotate, after which it decays away and is replaced by a helical current sheet. The neoclassical flow also generates an ion polarization current in the vicinity of the island chain which either increases or decreases the chain's radial width, depending on the direction of the flow. If the polarization effect is stabilizing, and exceeds a critical amplitude, then the helical island equilibrium becomes unstable, and the chain again decays away. The critical amplitude of the resonant harmonic of the external perturbation at which the island chain either unlocks or becomes unstable is calculated as a function of the pedestal ion pressure, the neoclassical poloidal and toroidal ion velocities and the poloidal and toroidal flow damping rates.

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Commercial tokamak reactors with resistive toroidal field magnets

International Nuclear Information System (INIS)

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

1984-01-01

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

OCLATOR (One Coil Low Aspect Toroidal Reactor)

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, S.

1980-02-01

A new approach to construct a tokamak-type reactor(s) is presented. Basically the return conductors of toroidal field coils are eliminated and the toroidal field coil is replaced by one single large coil, around which there will be placed several tokamaks or other toroidal devices. The elimination of return conductors should, in addition to other advantages, improve the accessibility and maintainability of the tokamaks and offer a possible alternative to the search for special materials to withstand large neutron wall loading, as the frequency of changeover would be increased due to minimum downtime. It also makes it possible to have a low aspect ratio tokamak which should improve the ..beta.. limit, so that a low toroidal magnetic field strength might be acceptable, meaning that the NbTi superconducting wire could be used. This system is named OCLATOR (One Coil Low Aspect Toroidal Reactor).

System design of toroidal field power supply of CDD tokamak

Energy Technology Data Exchange (ETDEWEB)

Liu, Zheng Zhi

1996-12-01

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

System design of toroidal field power supply of CDD tokamak

International Nuclear Information System (INIS)

Liu, Zheng Zhi.

1996-12-01

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

Ballooning instabilities in tokamaks with sheared toroidal flows

International Nuclear Information System (INIS)

Waelbroeck, F.L.; Chen, L.

1990-11-01

The stability of ballooning modes in the presence of sheared toroidal flows is investigated. The eigenmodes are shown to be related by a Fourier transformation to the non-exponentially growing Floquet solutions found by Cooper. It is further shown that the problem cannot be reduced further than to a two dimensional partial differential equation. Next, the generalized ballooning equation is solved analytically for a circular tokamak equilibrium with sonic flows, but with a small rotation shear compared to the sound speed. With this ordering, the centrifugal forces are comparable to the pressure gradient forces driving the instability, but coupling of the mode with the sound wave is avoided. A new stability criterion is derived which explicitly demonstrates that flow shear is stabilizing at constant centrifugal force gradient. 34 refs

OCLATOR (One Coil Low Aspect Toroidal Reactor)

International Nuclear Information System (INIS)

Yoshikawa, S.

1980-02-01

A new approach to construct a tokamak-type reactor(s) is presented. Basically the return conductors of toroidal field coils are eliminated and the toroidal field coil is replaced by one single large coil, around which there will be placed several tokamaks or other toroidal devices. The elimination of return conductors should, in addition to other advantages, improve the accessibility and maintainability of the tokamaks and offer a possible alternative to the search for special materials to withstand large neutron wall loading, as the frequency of changeover would be increased due to minimum downtime. It also makes it possible to have a low aspect ratio tokamak which should improve the β limit, so that a low toroidal magnetic field strength might be acceptable, meaning that the NbTi superconducting wire could be used. This system is named OCLATOR

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

International Nuclear Information System (INIS)

Kostek, C.A.

1983-01-01

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

Toroidal current asymmetry in tokamak disruptions

Science.gov (United States)

Strauss, H. R.

2014-10-01

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

The effect of sheared toroidal rotation on pressure driven magnetic islands in toroidal plasmas

Energy Technology Data Exchange (ETDEWEB)

Hegna, C. C. [Departments of Engineering Physics and Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States)

2016-05-15

The impact of sheared toroidal rotation on the evolution of pressure driven magnetic islands in tokamak plasmas is investigated using a resistive magnetohydrodynamics model augmented by a neoclassical Ohm's law. Particular attention is paid to the asymptotic matching data as the Mercier indices are altered in the presence of sheared flow. Analysis of the nonlinear island Grad-Shafranov equation shows that sheared flows tend to amplify the stabilizing pressure/curvature contribution to pressure driven islands in toroidal tokamaks relative to the island bootstrap current contribution. As such, sheared toroidal rotation tends to reduce saturated magnetic island widths.

Resistive toroidal stability of internal kink modes in circular and shaped tokamaks

International Nuclear Information System (INIS)

Bondeson, A.; Luetjens, H.; Vlad, G.

1991-12-01

The linear resistive magnetohydrodynamical (MHD) stability of the n=1 internal kink mode in tokamaks is studied by toroidal computations. The stabilizing influence of small aspect ratio is confirmed, but it is found that shaping of the cross section influences the internal kink mode significantly. For finite pressure and small resistivity, curvature effects at the q=1 surface make the stability sensitively dependent on shape, and ellipticity (including JET shape) is destabilizing. Only a very restricted set of finite pressure equilibria is completely stable for q 0 <1. A typical result is that the resistive kink mode is slowed down by toroidal effects to a weak tearing/resistive interchange mode. It is suggested that weak resistive instabilities are stabilized during the ramp phase of the sawteeth by effects not included in the linear resistive MHD model. Possible mechanisms for triggering a sawtooth crash are discussed. (author) 18 figs., 34 refs

Unstable universal drift eigenmodes in toroidal plasmas

International Nuclear Information System (INIS)

Cheng, C.Z.; Chen, L.

1980-01-01

The eigenmode equation describing ballooning collisionless drift instabilities is analyzed both analytically and numerically. A new branch of eigenmodes, which corresponds to quasi-bound states due to toroidal coupling effects such as ion delB drifts, is shown to be destabilized by electron Landau damping for typical tokamak parameters. This branch cannot be understood by the strong coupling approximation. However, the slab-like (Pearlstein--Berk-type) branch is found to remain stable and experience enhanced shear damping

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

International Nuclear Information System (INIS)

Hatch, A.J.

1975-01-01

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

Neoclassical offset toroidal velocity and auxiliary ion heating in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Lazzaro, E., E-mail: lazzaro@ifp.cnr.it [Istituto di Fisica del Plasma CNR (Italy)

2016-05-15

In conditions of ideal axisymmetry, for a magnetized plasma in a generic bounded domain, necessarily toroidal, the uniform absorption of external energy (e.g., RF or any isotropic auxiliary heating) cannot give rise to net forces or torques. Experimental evidence on contemporary tokamaks shows that the near central absorption of RF heating power (ICH and ECH) and current drive in presence of MHD activity drives a bulk plasma rotation in the co-I{sub p} direction, opposite to the initial one. Also the appearance of classical or neoclassical tearing modes provides a nonlinear magnetic braking that tends to clamp the rotation profile at the q-rational surfaces. The physical origin of the torque associated with P{sub RF} absorption could be due the effects of asymmetry in the equilibrium configuration or in power deposition, but here we point out also an effect of the response of the so-called neoclassical offset velocity to the power dependent heat flow increment. The neoclassical toroidal viscosity due to internal magnetic kink or tearing modes tends to relax the plasma rotation to this asymptotic speed, which in absence of auxiliary heating is of the order of the ion diamagnetic velocity. It can be shown by kinetic and fluid calculations, that the absorption of auxiliary power by ions modifies this offset proportionally to the injected power thereby forcing the plasma rotation in a direction opposite to the initial, to large values. The problem is discussed in the frame of the theoretical models of neoclassical toroidal viscosity.

Predictions of toroidal rotation and torque sources arising in non-axisymmetric perturbed magnetic fields in tokamaks

Science.gov (United States)

Honda, M.; Satake, S.; Suzuki, Y.; Shinohara, K.; Yoshida, M.; Narita, E.; Nakata, M.; Aiba, N.; Shiraishi, J.; Hayashi, N.; Matsunaga, G.; Matsuyama, A.; Ide, S.

2017-11-01

Capabilities of the integrated framework consisting of TOPICS, OFMC, VMEC and FORTEC-3D, have been extended to calculate toroidal rotation in fully non-axisymmetric perturbed magnetic fields for demonstrating operation scenarios in actual tokamak geometry and conditions. The toroidally localized perturbed fields due to the test blanket modules and the tangential neutral beam ports in ITER augment the neoclassical toroidal viscosity (NTV) substantially, while do not significantly influence losses of beam ions and alpha particles in an ITER L-mode discharge. The NTV takes up a large portion of total torque in ITER and fairly decelerates toroidal rotation, but the change in toroidal rotation may have limited effectiveness against turbulent heat transport. The error field correction coils installed in JT-60SA can externally apply the perturbed fields, which may alter the NTV and the resultant toroidal rotation profiles. However, the non-resonant n=18 components of the magnetic fields arising from the toroidal field ripple mainly contribute to the NTV, regardless of the presence of the applied field by the coil current of 10 kA , where n is the toroidal mode number. The theoretical model of the intrinsic torque due to the fluctuation-induced residual stress is calibrated by the JT-60U data. For five JT-60U discharges, the sign of the calibration factor conformed to the gyrokinetic linear stability analysis and a range of the amplitude thereof was revealed. This semi-empirical approach opens up access to an attempt on predicting toroidal rotation in H-mode plasmas.

Calculation of modification to the toroidal magnetic field of the Tokamak Novillo. Part II; Calculo de modificacion al campo magnetico toroidal del Tokamak nivillo. Parte II

Energy Technology Data Exchange (ETDEWEB)

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

1992-03-15

In a cylindrical magnetic topology. the confined plasma experiences 'classic' collisional transport phenomena. When bending the cylinder with the purpose of forming a toro, the magnetic field that before was uniform now it has a radial gradient which produces an unbalance in the magnetic pressure that is exercised on the plasma in the transverse section of the toro. This gives place to transport phenomena call 'neo-classicist'. In this work the structure of the toroidal magnetic field produced by toroidal coils of triangular form, to which are added even of coils of compensation with form of half moon is analyzed. With this type of coils it is looked for to minimize the radial gradient of the toroidal magnetic field. The values and characteristics of B (magnetic field) in perpendicular planes to the toro in different angular positions in the toroidal direction, looking for to cover all the cases of importance are exhibited. (Author)

Toroidal rotation studies in KSTAR

Science.gov (United States)

Lee, S. G.; Lee, H. H.; Yoo, J. W.; Kim, Y. S.; Ko, W. H.; Terzolo, L.; Bitter, M.; Hill, K.; KSTAR Team

2014-10-01

Investigation of the toroidal rotation is one of the most important topics for the magnetically confined fusion plasma researches since it is essential for the stabilization of resistive wall modes and its shear plays an important role to improve plasma confinement by suppressing turbulent transport. The most advantage of KSTAR tokamak for toroidal rotation studies is that it equips two main diagnostics including the high-resolution X-ray imaging crystal spectrometer (XICS) and charge exchange spectroscopy (CES). Simultaneous core toroidal rotation and ion temperature measurements of different impurity species from the XICS and CES have shown in reasonable agreement with various plasma discharges in KSTAR. It has been observed that the toroidal rotation in KSTAR is faster than that of other tokamak devices with similar machine size and momentum input. This may due to an intrinsically low toroidal field ripple and error field of the KSTAR device. A strong braking of the toroidal rotation by the n = 1 non-resonant magnetic perturbations (NRMPs) also indicates these low toroidal field ripple and error field. Recently, it has been found that n = 2 NRMPs can also damp the toroidal rotation in KSTAR. The detail toroidal rotation studies will be presented. Work supported by the Korea Ministry of Science, ICT and Future Planning under the KSTAR project.

Steady-state resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1979-12-01

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

Prospects for toroidal fusion reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.D.

1994-01-01

Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate components in engineering, technology, and unit costs, should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, ∼2000 MW(e), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges

Experimental studies of plasma confinement in toroidal systems

International Nuclear Information System (INIS)

Bodin, H.A.B.; Keen, B.E.

1977-01-01

In this article the closed-line magnetic field approach to the plasma isolation and confinement problem in toroidal systems is reviewed. The theoretical aspects of closed-line magnetic field systems, indicating that topologically such systems are toroidal, are surveyed under the headings; topology of closed-line systems, equilibrium in different configurations and classification of toroidal devices, MHD stability, non-ideal effects in MHD stability, microscopic stability, and plasma energy loss. A section covering the experimental results of plasma confinement in toroidal geometry considers Stellerators, Tokamaks, toroidal pinch -the reversed-field pinch, screw pinches and high-β Tokamaks, Levitrons and multipoles (internal-ring devices), and miscellaneous toroidal containment devices. Recent achievements and the present position are discussed with reference to the status of Tokamak research, low-β stellerator research and high-β research. It is concluded from the continuing progress made in this research that the criteria for the magnetic containment of plasmas can be met. Further, it is concluded that the construction of a successful and economic fusion reactor is within the scope of advancing science and technology. 250 references. (U.K.)

Experimental studies of plasma confinement in toroidal systems

Energy Technology Data Exchange (ETDEWEB)

Bodin, H A.B.; Keen, B E [UKAEA, Abingdon. Culham Lab.

1977-12-01

In this article the closed-line magnetic field approach to the plasma isolation and confinement problem in toroidal systems is reviewed. The theoretical aspects of closed-line magnetic field systems, indicating that topologically such systems are toroidal, are surveyed under the headings; topology of closed-line systems, equilibrium in different configurations and classification of toroidal devices, MHD stability, non-ideal effects in MHD stability, microscopic stability, and plasma energy loss. A section covering the experimental results of plasma confinement in toroidal geometry considers Stellerators, Tokamaks, toroidal pinch -the reversed-field pinch, screw pinches and high-..beta.. Tokamaks, Levitrons and multipoles (internal-ring devices), and miscellaneous toroidal containment devices. Recent achievements and the present position are discussed with reference to the status of Tokamak research, low-..beta.. stellerator research and high-..beta.. research. It is concluded from the continuing progress made in this research that the criteria for the magnetic containment of plasmas can be met. Further, it is concluded that the construction of a successful and economic fusion reactor is within the scope of advancing science and technology. 250 references.

Advanced Tokamak Stability Theory

Science.gov (United States)

Zheng, Linjin

2015-03-01

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

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

Science.gov (United States)

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

2007-08-10

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

Effects of a sheared toroidal rotation on the stability boundary of the MHD modes in the tokamak edge pedestal

International Nuclear Information System (INIS)

Aiba, N.; Tokuda, S.; Oyama, N.; Ozeki, T.; Furukawa, M.

2009-01-01

Effects of a sheared toroidal rotation are investigated numerically on the stability of the MHD modes in the tokamak edge pedestal, which relate to the type-I edge-localized mode. A linear MHD stability code MINERVA is newly developed for solving the Frieman-Rotenberg equation that is the linear ideal MHD equation with flow. Numerical stability analyses with this code reveal that the sheared toroidal rotation destabilizes edge localized MHD modes for rotation frequencies which are experimentally achievable, though the ballooning mode stability changes little by rotation. This rotation effect on the edge MHD stability becomes stronger as the toroidal mode number of the unstable MHD mode increases when the stability analysis was performed for MHD modes with toroidal mode numbers smaller than 40. The toroidal mode number of the unstable MHD mode depends on the stabilization of the current-driven mode and the ballooning mode by increasing the safety factor. This dependence of the toroidal mode number of the unstable mode on the safety factor is considered to be the reason that the destabilization by toroidal rotation is stronger for smaller edge safety factors.

A simple toroidal shell model for the study of feedback stabilization of resistive wall modes in a tokamak plasma

International Nuclear Information System (INIS)

Jhang, Hogun

2008-01-01

A study is conducted on the feedback stabilization of resistive wall modes (RWMs) in a tokamak plasma using a toroidal shell model. An analytically tractable form of the RWM dispersion relation is derived in the presence of a set of discrete feedback coil currents. A parametric study is carried out to optimize the feedback system configuration. It is shown that the total toroidal angle of a resistive wall spanned by the feedback coils and the poloidal angular extent of a feedback coil are crucial parameters to determine the efficacy of the feedback system

Theory of high-n toroidicity-induced shear Alfven eigenmode in tokamaks

International Nuclear Information System (INIS)

Fu, G.Y.; Cheng, C.Z.; Princeton Univ., NJ

1989-07-01

High-n WKB-ballooning mode equation is employed to study toroidicity-induced shear Alfven eigenmodes (TAE) in the δ - α space, where δ = (r/q)(dq/dr) is the magnetic shear, and α = -(2Rq 2 /B 2 )(dp/dr) is the normalized pressure gradient for tokamak plasmas. In the ballooning mode first stability region, TAE modes are found to exist only for α less than some critical value α c . We also find that these TAE modes reappear in the ballooning mode second stability region for bands of α values. The global envelope structures of these TAE modes are studied by WKB method and are found to be bounded radially if the local mode frequency has a maximum in radius. 15 refs., 14 figs

Evaluation of toroidal torque by non-resonant magnetic perturbations in tokamaks for resonant transport regimes using a Hamiltonian approach

Energy Technology Data Exchange (ETDEWEB)

Albert, Christopher G.; Heyn, Martin F.; Kapper, Gernot; Kernbichler, Winfried; Martitsch, Andreas F. [Fusion@ÖAW, Institut für Theoretische Physik - Computational Physics, Technische Universität Graz, Petersgasse 16, 8010 Graz (Austria); Kasilov, Sergei V. [Fusion@ÖAW, Institut für Theoretische Physik - Computational Physics, Technische Universität Graz, Petersgasse 16, 8010 Graz (Austria); Institute of Plasma Physics, National Science Center “Kharkov Institute of Physics and Technology,” ul. Akademicheskaya 1, 61108 Kharkov (Ukraine)

2016-08-15

Toroidal torque generated by neoclassical viscosity caused by external non-resonant, non-axisymmetric perturbations has a significant influence on toroidal plasma rotation in tokamaks. In this article, a derivation for the expressions of toroidal torque and radial transport in resonant regimes is provided within quasilinear theory in canonical action-angle variables. The proposed approach treats all low-collisional quasilinear resonant neoclassical toroidal viscosity regimes including superbanana-plateau and drift-orbit resonances in a unified way and allows for magnetic drift in all regimes. It is valid for perturbations on toroidally symmetric flux surfaces of the unperturbed equilibrium without specific assumptions on geometry or aspect ratio. The resulting expressions are shown to match the existing analytical results in the large aspect ratio limit. Numerical results from the newly developed code NEO-RT are compared to calculations by the quasilinear version of the code NEO-2 at low collisionalities. The importance of the magnetic shear term in the magnetic drift frequency and a significant effect of the magnetic drift on drift-orbit resonances are demonstrated.

Neoclassical transport in toroidal systems

International Nuclear Information System (INIS)

Wobig, H.

1992-01-01

The neoclassical theory of general toroidal equilibria is reformulated. The toroidal equilibrium of tokamaks and stellarators are described in Hamada coordinates. The relevant geometrical parameters are identified and it is shown how the reduction of Pfirsch-Schluter currents affects neoclassical transport and bootstrap effects. General flux-friction relations between thermodynamic forces and fluxes are derived. In drift-kinetic approximation the neoclassical transport coefficients are Onsager symmetric. Since a toroidal loop voltage is included, the theory is valid for all toroidal systems. (Author)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

2014-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-09-01

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

Magnetic field structure of experimental high beta tokamak equilibria

International Nuclear Information System (INIS)

Deniz, A.V.

1986-01-01

The magnetic field structure of several low and high β tokamaks in the Columbia High Beta Tokamak (HBT) was determined by high-impedance internal magnetic probes. From the measurement of the magnetic field, the poloidal flux, toroidal flux, toroidal current, and safety factor are calculated. In addition, the plasma position and cross-sectional shape are determined. The extent of the perturbation of the plasma by the probe was investigated and was found to be acceptably small. The tokamaks have major radii of approx.0.24 m, minor radii of approx.0.05 m, toroidal plasma current densities of approx.10 6 A/m 2 , and line-integrated electron densities of approx.10 20 m -2 . The major difference between the low and high β tokamaks is that the high β tokamak was observed to have an outward shift in major radius of both the magnetic center and peak of the toroidal current density. The magnetic center moves inward in major radius after 20 to 30 μsec, presumably because the plasma maintains major radial equilibrium as its pressure decreases from radiation due to impurity atoms. Both the equilibrium and the production of these tokamaks from a toroidal field stabilized z-pinch are modeled computationally. One tokamak evolves from a state with low β features, through a possibly unstable state, to a state with high β features

Effects of toroidal field ripple on suprathermal ions in tokamak plasmas

International Nuclear Information System (INIS)

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

1980-02-01

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

Effects of toroidal field ripple on suprathermal ions in tokamak plasmas

International Nuclear Information System (INIS)

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

1981-01-01

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

Design study of toroidal magnets for tokamak experimental power reactors

International Nuclear Information System (INIS)

Stekly, Z.J.J.; Lucas, E.J.

1976-12-01

This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed

PPPL tokamak program

International Nuclear Information System (INIS)

Furth, H.P.

1984-10-01

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

Finite element and node point generation computer programs used for the design of toroidal field coils in tokamak fusion devices

International Nuclear Information System (INIS)

Smith, R.A.

1975-06-01

The structural analysis of toroidal field coils in Tokamak fusion machines can be performed with the finite element method. This technique has been employed for design evaluations of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The application of the finite element method can be simplified with computer programs that are used to generate the input data for the finite element code. There are three areas of data input where significant automation can be provided by supplementary computer codes. These concern the definition of geometry by a node point mesh, the definition of the finite elements from the geometric node points, and the definition of the node point force/displacement boundary conditions. The node point forces in a model of a toroidal field coil are computed from the vector cross product of the coil current and the magnetic field. The computer programs named PDXNODE and ELEMENT are described. The program PDXNODE generates the geometric node points of a finite element model for a toroidal field coil. The program ELEMENT defines the finite elements of the model from the node points and from material property considerations. The program descriptions include input requirements, the output, the program logic, the methods of generating complex geometries with multiple runs, computational time and computer compatibility. The output format of PDXNODE and ELEMENT make them compatible with PDXFORC and two general purpose finite element computer codes: (ANSYS) the Engineering Analysis System written by the Swanson Analysis Systems, Inc., and (WECAN) the Westinghouse Electric Computer Analysis general purpose finite element program. The Fortran listings of PDXNODE and ELEMENT are provided

Analytical modelling of resistive wall mode stabilization by rotation in toroidal tokamak plasmas

International Nuclear Information System (INIS)

Ham, C J; Gimblett, C G; Hastie, R J

2011-01-01

Stabilization of the resitive wall mode (RWM) may allow fusion power to be doubled for a given magnetic field in advanced tokamak operation. Experimental evidence from DIII-D and other machines suggests that plasma rotation can stabilize the RWM. Several authors (Finn 1995 Phys. Plasmas 2 3782, Bondeson and Xie 1997 Phys. Plasmas 4 2081) have constructed analytical cylindrical models for the RWM, but these do not deal with toroidal effects. The framework of Connor et al (1988 Phys. Fluids 31 577) is used to develop ideal plasma analytic models with toroidicity included. Stepped pressure profiles and careful ordering of terms are used to simplify the analysis. First, a current driven kink mode model is developed and a dispersion relation for arbitrary current profile is calculated. Second, the external pressure driven kink mode is similarly investigated as the most important RWM arises from this mode. Using this latter model it is found that the RWM is stabilized by Alfven continuum damping with rotation levels similar to those seen in experiments. An expression for the stability of the external kink mode for more general current profiles and a resistive wall is derived in the appendix.

Modular tokamak magnetic system

International Nuclear Information System (INIS)

Yang, T.F.

1988-01-01

This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

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

International Nuclear Information System (INIS)

Peysson, Y.; Decker, J.

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1976-10-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

A Toroidally Symmetric Plasma Simulation code for design of position and shape control on tokamak plasmas

International Nuclear Information System (INIS)

Takase, Haruhiko; Senda, Ikuo

1999-01-01

A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)

Evaluation of the toroidal torque driven by external non-resonant non-axisymmetric magnetic field perturbations in a tokamak

Energy Technology Data Exchange (ETDEWEB)

Kasilov, Sergei V. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Institute of Plasma Physics National Science Center “Kharkov Institute of Physics and Technology” ul. Akademicheskaya 1, 61108 Kharkov (Ukraine); Kernbichler, Winfried; Martitsch, Andreas F.; Heyn, Martin F. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Maassberg, Henning [Max-Planck Institut für Plasmaphysik, D-17491 Greifswald (Germany)

2014-09-15

The toroidal torque driven by external non-resonant magnetic perturbations (neoclassical toroidal viscosity) is an important momentum source affecting the toroidal plasma rotation in tokamaks. The well-known force-flux relation directly links this torque to the non-ambipolar neoclassical particle fluxes arising due to the violation of the toroidal symmetry of the magnetic field. Here, a quasilinear approach for the numerical computation of these fluxes is described, which reduces the dimension of a standard neoclassical transport problem by one without model simplifications of the linearized drift kinetic equation. The only limiting condition is that the non-axisymmetric perturbation field is small enough such that the effect of the perturbation field on particle motion within the flux surface is negligible. Therefore, in addition to most of the transport regimes described by the banana (bounce averaged) kinetic equation also such regimes as, e.g., ripple-plateau and resonant diffusion regimes are naturally included in this approach. Based on this approach, a quasilinear version of the code NEO-2 [W. Kernbichler et al., Plasma Fusion Res. 3, S1061 (2008).] has been developed and benchmarked against a few analytical and numerical models. Results from NEO-2 stay in good agreement with results from these models in their pertinent range of validity.

Toroidal field ripple effects in large tokamaks

International Nuclear Information System (INIS)

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

1975-01-01

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

EFIT tokamak equilibria with toroidal flow and anisotropic pressure using the two-temperature guiding-centre plasma

International Nuclear Information System (INIS)

Fitzgerald, M.; Hole, M.J.; Appel, L.C.

2013-01-01

A new force balance model for the EFIT magnetohydrodynamic equilibrium technique for tokamaks is presented which includes the full toroidal flow and anisotropy changes to the Grad–Shafranov equation. The free functions are poloidal flux functions and all non-linear contributions to the toroidal current density are treated iteratively. The parallel heat flow approximation chosen for the model is that parallel temperature is a flux function and that both parallel and perpendicular pressures may be described using parallel and perpendicular temperatures. This choice for the fluid thermodynamics has been shown elsewhere to be the same as a guiding-centre kinetic solution of the same problem under the same assumptions. The model reduces identically to the static and isotropic Grad–Shafranov equation in the appropriate limit as different flux functions are set to zero. An analytical solution based on a modified Soloviev solution for non-zero toroidal flow and anisotropy is also presented. The force balance model has been demonstrated in the code EFIT TENSOR, a branch of the existing code EFIT++. Benchmark results for EFIT TENSOR are presented and the more complicated force balance model is found to converge to force balance similarly to the usual EFIT model and with comparable speed. (paper)

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

International Nuclear Information System (INIS)

Horie, T.

1987-07-01

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

Current control necessary for toroidal plasma equilibrium

International Nuclear Information System (INIS)

Nagao, S.

1987-01-01

It is shown that a significant amount of dipole current is necessary for the plasma equilibrium of toroidal configurations in general. Through the vector product with the poloidal field, this dipole current force has to balance with the hoop force of plasma pressure itself of the annular shape. The measurement of such a current of dipole type may be interesting for the confirmation of the plasma equilibrium in the toroidal system. Moreover it is certained that there is a new mode of a tokamak operation with such a dipole current component and with smaller vertical field than that based on the classical tokamak theory. (author) [pt

Supporting device for Toroidal coils

International Nuclear Information System (INIS)

Araki, Takao.

1985-01-01

Purpose: To reduce the response of a toroidal coil supporting device upon earthquakes and improve the earthquake proofness in a tokamak type thermonuclear device. Constitution: Structural materials having large longitudinal modulus and enduring great stresses, for example, stainless steels are used as the toroidal coil supporting legs and heat insulating structural materials are embedded in a nuclear reactor base mats below the supporting legs. Furthermore, heat insulating concretes are spiked around the heat insulating structural materials to prevent the intrusion of heat to the toroidal coils. The toroidal coils are kept at cryogenic state and superconductive state for the conductors. In this way, the period of proper vibrations of the toroidal coils and the toroidal coil supporting structures can be shortened thereby decreasing the seismic response. Furthermore, since the strength of the supporting legs is increased, the earthquake proofness of the coils can be improved. (Kamimura, M.)

Neoclassical toroidal viscosity calculations in tokamaks using a δf Monte Carlo simulation and their verifications.

Science.gov (United States)

Satake, S; Park, J-K; Sugama, H; Kanno, R

2011-07-29

Neoclassical toroidal viscosities (NTVs) in tokamaks are investigated using a δf Monte Carlo simulation, and are successfully verified with a combined analytic theory over a wide range of collisionality. A Monte Carlo simulation has been required in the study of NTV since the complexities in guiding-center orbits of particles and their collisions cannot be fully investigated by any means of analytic theories alone. Results yielded the details of the complex NTV dependency on particle precessions and collisions, which were predicted roughly in a combined analytic theory. Both numerical and analytic methods can be utilized and extended based on these successful verifications.

Effects of 3D Magnetic Perturbations on Toroidal Plasmas

International Nuclear Information System (INIS)

Callen, J.D.

2010-01-01

Full text: To lowest order tokamaks are two-dimensional (2D) axisymmetric magnetic systems. But small 3D magnetic perturbations (both externally applied and from plasma instabilities) have many interesting and useful effects on tokamak (and quasi-symmetric stellarator) plasmas. Plasma transport equations that include these effects, especially on diamagnetic-level toroidal plasma rotation, have recently been developed. The 3D magnetic perturbations and their plasma effects can be classified according to their toroidal mode number n: low n (1 to 5) resonant (q = m/n in plasma) and non-resonant fields, medium n (due to toroidal field ripple), and high n (due to microturbulence). This paper concentrates on low and medium n perturbations. Low n non-resonant magnetic fields induce a neoclassical toroidal viscosity (NTV) that damps toroidal plasma rotation throughout the plasma toward an offset flow in the counter-I p direction; recent tokamak experiments have confirmed and exploited these predictions by applying external low n non-resonant magnetic perturbations. Medium n perturbations have similar effects plus possible ripple trapping and resultant edge ion losses. A low n resonant magnetic field induces a toroidal plasma torque in the vicinity of the rational surface; when large enough it can stop plasma rotation there and lead to a locked mode, which often causes a plasma disruption. Externally applied 3D magnetic perturbations usually have many components; in the plasma their lowest n components are amplified by plasma responses, particularly at high beta. Low n plasma instabilities (e.g., NTMs, RWMs) cause additional 3D magnetic perturbations in tokamak plasmas; tearing modes can bifurcate the topology and form magnetic islands. Finally, multiple resonant magnetic perturbations (RMPs) can cause local magnetic stochasticity and influence H-mode edge pedestal transport. These various effects of 3D magnetic perturbations can be used to control the toroidal plasma

Superconducting magnets for toroidal fusion reactors

International Nuclear Information System (INIS)

Haubenreich, P.N.

1980-01-01

Fusion reactors will soon be employing superconducting magnets to confine plasma in which deuterium and tritium (D-T) are fused to produce usable energy. At present there is one small confinement experiment with superconducting toroidal field (TF) coils: Tokamak 7 (T-7), in the USSR, which operates at 4 T. By 1983, six different 2.5 x 3.5-m D-shaped coils from six manufacturers in four countries will be assembled in a toroidal array in the Large Coil Test Facility (LCTF) at Oak Ridge National Laboratory (ORNL) for testing at fields up to 8 T. Soon afterwards ELMO Bumpy Torus (EBT-P) will begin operation at Oak Ridge with superconducting TF coils. At the same time there will be tokamaks with superconducting TF coils 2 to 3 m in diameter in the USSR and France. Toroidal field strength in these machines will range from 6 to 9 T. NbTi and Nb 3 Sn, bath cooling and forced flow, cryostable and metastable - various designs are being tried in this period when this new application of superconductivity is growing and maturing

Toroidal simulation magnet tests

International Nuclear Information System (INIS)

Walstrom, P.L.; Domm, T.C.

1975-01-01

A number of different schemes for testing superconducting coils in a simulated tokamak environment are analyzed for their merits relative to a set of test criteria. Two of the concepts are examined in more detail: the so-called cluster test scheme, which employs two large background field coils, one on either side of the test coil, and the compact torus, a low-aspect ratio toroidal array of a small number of coils in which all of the coils are essentially test coils. Simulation of the pulsed fields of the tokamak is discussed briefly

Computer simulation of transport driven current in tokamaks

International Nuclear Information System (INIS)

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

1993-01-01

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

Creating poloidal flux in a tokamak plasma with low frequency waves

International Nuclear Information System (INIS)

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

1993-01-01

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

The combined toroidicity, ellipticity and triangularity effects on the energy deposition of Alfven modes in pre-heated, low aspect ratio tokamaks

Energy Technology Data Exchange (ETDEWEB)

Cuperman, S. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel); Bruma, C. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel) and College of Judea and Samaria, 44837 Ariel (Israel)]. E-mail: edycb@post.tau.ac.il; Komoshvili, K. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel); College of Judea and Samaria, 44837 Ariel (Israel)

2007-03-05

The combined plasma non-uniformity effects on the energy deposition of Alfven waves launched by an external antenna in pre-heated spherical tokamaks are investigated. The following relevant physical processes are here possible: (a) the emergence of gaps in the shear Alfven continuum spectrum and the generation of discrete global Alfven eigenmodes with frequencies inside the gaps; (b) multi-wave interactions, interactions of gaps of the same kind (e.g., toroidicity induced) and of different kinds (toroidicity, ellipticity and triangularity induced) as well as of secondary order gaps arising when a pair of modes is coupled to one or more modes through other coupling parameters; (c) basic wave-plasma interactions as propagation, reflection, mode-conversion, tunneling and deposition. Thus, we solved numerically the full 2D wave equations for the vector and scalar potentials, using a quite general two-fluid resistive tensor-operator, without any geometrical limitations. The results obtained indicate the existence of antenna-launched wave characteristics for which the power is most efficiently coupled in outer regions of plasmas, which is of special interest for low aspect ratio tokamaks, e.g., for the generation of non-inductive current drive as well as for turbulence suppression and transport barriers formation.

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

Energy Technology Data Exchange (ETDEWEB)

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

1966-04-15

la colonne, qui peut s'expliquer par la presence, a l'interieur de la colonne, de flux de derive d'ions ayant de grandes vitesses transversales. (author) [Spanish] En la instalacion Tokamak-3 el plasma de hidrogeno se obtiene y calienta con una corriente electrica anular de 40 a 60 kA. El tiempo de paso de la corriente es de 20 a 30 ms. Muchos de los experimentos se efectuaron con un campo magnetico longitudinal estabilizado de 25 kOe. La componente transversal del campo magnetico disperso se compenso con ayuda de corrientes procedentes de circuitos correctores especiales. Durante la descarga no tuvo lugar ningun desplazamiento perceptible del centro de la columna con respecto al plano ecuatorial del toro y se observo una deriva de la columna de plasma 'hacia el exterior' (aumento del radio mayor del circuito). Este movimiento puede ser provocado por un cambio del radio de la columna de corriente, por el calentamiento del plasma o por el amortiguamiento de las corrientes de Foucault en el recipiente conductor. En estas condiciones, se consiguio obtener en la instalacion Tokamak-3 una columna de plasma macroscopicamente estable; sin embargo, la temperatura del plasma fue inferior a la que era de esperar en ausencia de perdidas anomalas de energia. Los procesos que tienen lugar durante la interaccion de la columna plasmatica con el diafragma originan grandes perdidas de energfa. Esta interaccion pudo atenuarse dando al diafragma una forma especial y utilizando la propiedad de la columna de desplazarse hacia el exterior en el transcurso del proceso. El problema investigado consistia en determinar varios parametros del plasma en estas condiciones, en particular las temperaturas electronica y ionica. Para determinar la temperatura electronica basandose en los cambios de la resistencia del plasma, es preciso conocer ademas de las caracteristicas electricas de la columna, la ley que rige el cambio del radio de esta ultima en funcion del tiempo. La solucion, utilizando un

Toroidal Plasma Thruster for Interplanetary and Interstellar Space Flights

International Nuclear Information System (INIS)

Gorelenkov, N.N.; Zakharov, L.E.; Gorelenkova, M.V.

2001-01-01

This work involves a conceptual assessment for using the toroidal fusion reactor for deep space interplanetary and interstellar missions. Toroidal thermonuclear fusion reactors, such as tokamaks and stellarators, are unique for space propulsion, allowing for a design with the magnetic configuration localized inside toroidal magnetic field coils. Plasma energetic ions, including charged fusion products, can escape such a closed configuration at certain conditions, a result of the vertical drift in toroidal rippled magnetic field. Escaping particles can be used for direct propulsion (since toroidal drift is directed one way vertically) or to create and heat externally confined plasma, so that the latter can be used for propulsion. Deuterium-tritium fusion neutrons with an energy of 14.1 MeV also can be used for direct propulsion. A special design allows neutrons to escape the shield and the blanket of the tokamak. This provides a direct (partial) conversion of the fusion energy into the directed motion of the propellant. In contrast to other fusion concepts proposed for space propulsion, this concept utilizes the natural drift motion of charged particles out of the closed magnetic field configuration

Current drive by spheromak injection into a tokamak

International Nuclear Information System (INIS)

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

1990-01-01

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

Internal (m=1, n=1) and (m=2, n=1) resistive modes in the toroidal tokamak with circular cross-section

International Nuclear Information System (INIS)

Bussac, M.N.; Pellat, R.; Edery, D.; Soule, J.L.

1977-01-01

A linear analysis is presented of the toroidal coupling between the internal resistive modes (m=1, n=1) and (m=2, n=1) in the tokamak with circular cross-section. The resistive and diamagnetic effects are included in the singular layers where the safety factor q takes respectively the values one and two. By expanding the MHD equations in powers of epsilon, the local inverse of the aspect ratio, a system of two coupled equations is obtained for the harmonic amplitudes. When the shear is finite on q=1 the toroidal coupling is negligible. In the opposite limit, one can explain (a) the experimental behaviour of the (m=1, n=1) mode before the internal disruption, and (b) the simultaneous observation of the modes (m=1, n=1) and (m=2, n=1) before the main disruption. (author)

Qualifying tests for TRIAM-1M superconducting toroidal magnetic field coil

Energy Technology Data Exchange (ETDEWEB)

Nakanura, Yukio; Hiraki, Naoji; Nakamura, Kazuo; Tanaka, Masayoshi; Nagao, Akihiro; Kawasaki, Shoji; Itoh, Satoshi

1984-09-01

In the strong toroidal magnetic field experimental facility ''TRIAM-1M'' currently under construction, construction of the superconducting toroidal magnetic field coil and the following qualifying tests conducted on the full-scale superconducting toroidal magnetic field coil actually fabricated are described: (1) coil excitation test, (2) superconducting stability test, (3) external magnetic field application test, and (4) high-speed excitation test. On the basis of these test results, stability was evaluated of the superconducting coil being operated in the tokamak device. In normal tokamak operation, there occurs no normal conduction transition. At the time of plasma disruption, though this transition takes place in part of the coil, the superconducting state is immediately restored. By its electromagnetic force analysis, the superconducting coil is also stable in structure.

Progress in gyrokinetic simulations of toroidal ITG turbulence

International Nuclear Information System (INIS)

Nevins, W.M.; Dimits, A.M.; Cohen, B.I.; Shumaker, D.E.

2001-01-01

The 3-D nonlinear toroidal gyrokinetic simulation code PG3EQ is used to study toroidal ion temperature gradient (ITG) driven turbulence - a key cause of the anomalous transport that limits tokamak plasma performance. Systematic studies of the dependence of ion thermal transport on various parameters and effects are presented, including dependence on E-vectorxB-vector and toroidal velocity shear, sensitivity to the force balance in simulations with radial temperature gradient variation, and the dependences on magnetic shear and ion temperature gradient. (author)

Prandtl number of toroidal plasmas

International Nuclear Information System (INIS)

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

1993-06-01

Theory of the L-mode confinement in toroidal plasmas is developed. The Prandtl number, the ratio between the ion viscosity and the thermal conductivity is obtained for the anomalous transport process which is caused by the self-sustained turbulence in the toroidal plasma. It is found that the Prandtl number is of order unity both for the ballooning mode turbulence in tokamaks and for the interchange mode turbulence in helical system. The influence on the anomalous transport and fluctuation level is evaluated. Hartmann number and magnetic Prandtl number are also discussed. (author)

Compact toroid injection system for JFT-2M

Energy Technology Data Exchange (ETDEWEB)

Fukumoto, N. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan)]. E-mail: fukumotn@eng.u-hyogo.ac.jp; Ogawa, H. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Nagata, M. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Uyama, T. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Shibata, T. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Kashiwa, Y. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Suzuki, S. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Kusama, Y. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2006-11-15

The compact toroid (CT) injection system for JFT-2M is composed of a CT injector, a gas delivery and vacuum system, a power supply system, and a diagnostics system. In particular, the power supply system delivers high performance for CT formation and acceleration. The CT formation capacitor bank unit achieved a formation current of 350 kA with a rise time less than 10 {mu}s. Although the CT acceleration bank units are equipped with 14 ignitron switches instead of gap switches to attenuate the discharge noise level, an acceleration current of 400 kA with a short rise time of 9 {mu}s is controlled within a jitter of much less than 1 {mu}s. The resulting CT velocity and mass density satisfy the requirements for CT penetration into the tokamak plasma core at a toroidal field of 1 T. This CT injection system is thus suitable for CT injection in a middle-sized tokamak plasma such as the JFT-2M tokamak.

Compact toroid injection system for JFT-2M

International Nuclear Information System (INIS)

Fukumoto, N.; Ogawa, H.; Nagata, M.; Uyama, T.; Shibata, T.; Kashiwa, Y.; Suzuki, S.; Kusama, Y.

2006-01-01

The compact toroid (CT) injection system for JFT-2M is composed of a CT injector, a gas delivery and vacuum system, a power supply system, and a diagnostics system. In particular, the power supply system delivers high performance for CT formation and acceleration. The CT formation capacitor bank unit achieved a formation current of 350 kA with a rise time less than 10 μs. Although the CT acceleration bank units are equipped with 14 ignitron switches instead of gap switches to attenuate the discharge noise level, an acceleration current of 400 kA with a short rise time of 9 μs is controlled within a jitter of much less than 1 μs. The resulting CT velocity and mass density satisfy the requirements for CT penetration into the tokamak plasma core at a toroidal field of 1 T. This CT injection system is thus suitable for CT injection in a middle-sized tokamak plasma such as the JFT-2M tokamak

Long-wavelength microinstabilities in toroidal plasmas

International Nuclear Information System (INIS)

Tang, W.M.; Rewoldt, G.

1993-01-01

Realistic kinetic toroidal eigenmode calculations have been carried out to support a proper assessment of the influence of long-wavelength microturbulence on transport in tokamak plasmas. In order to efficiently evaluate large-scale kinetic behavior extending over many rational surfaces, significant improvements have been made to a toroidal finite element code used to analyze the fully two-dimensional (r,θ) mode structures of trapped-ion and toroidal ion temperature gradient (ITG) instabilities. It is found that even at very long wavelengths, these eigenmodes exhibit a strong ballooning character with the associated radial structure relatively insensitive to ion Landau damping at the rational surfaces. In contrast to the long-accepted picture that the radial extent of trapped-ion instabilities is characterized by the ion-gyroradius-scale associated with strong localization between adjacent rational surfaces, present results demonstrate that under realistic conditions, the actual scale is governed by the large-scale variations in the equilibrium gradients. Applications to recent measurements of fluctuation properties in Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. (International Atomic Energy Agency, Vienna, 1985), Vol. 1, p. 29] L-mode plasmas indicate that the theoretical trends appear consistent with spectral characteristics as well as rough heuristic estimates of the transport level. Benchmarking calculations in support of the development of a three-dimensional toroidal gyrokinetic code indicate reasonable agreement with respect to both the properties of the eigenfunctions and the magnitude of the eigenvalues during the linear phase of the simulations of toroidal ITG instabilities

Unstable universal drift eigenmodes in toroidal plasmas

International Nuclear Information System (INIS)

Cheng, C.Z.; Chen, L.

1979-08-01

The eigenmode equation describing ballooning collisionless drift instabilities is analyzed both analytically and numerically. A new branch of eigenmodes, which corresponds to quasi-bound states due to the finite toroidicity, is shown to be destabilized by electron Landau damping for typical Tokamak parameters. This branch cannot be understood by the strong coupling approximation. However, the slab-like (Pearlstein-Berk type) branch is found to remain stable and experience enhanced shear damping due to finite toroidicity

Continuous tokamaks

International Nuclear Information System (INIS)

Peng, Y.K.M.

1978-04-01

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

Empirical scaling for present Ohmically heated tokamaks

International Nuclear Information System (INIS)

Daughney, C.

1975-01-01

Experimental results from the Adiabatic Toroidal Compressor (ATC) tokamak are used to obtain empirical scaling laws for the average electron temperature and electron energy confinement time as functions of the average electron density, the effective ion charge, and the plasma current. These scaling laws are extended to include dependence upon minor and major plasma radius and toroidal field strength through a comparison of the various tokamaks described in the literature. Electron thermal conductivity is the dominant loss process for the ATC tokamak. The parametric dependences of the observed electron thermal conductivity are not explained by present theoretical considerations. The electron temperature obtained with Ohmic heating is shown to be a function of current density - which will not be increased in the next generation of large tokamaks. However, the temperature dependence of the electron energy confinement time suggests that significant improvement in confinement time will be obtained with supplementary electron heating. (author)

Compact tokamak reactors. Part 1 (analytic results)

International Nuclear Information System (INIS)

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

1996-01-01

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

Particle simulations in toroidal geometry

International Nuclear Information System (INIS)

Aydemir, A.Y.

1992-09-01

A computational tool to be used in kinetic simulations of toroidal plasmas is being developed. The initial goal of the project is to develop an electrostatic gyrokinetic model for studying transport and stability problems in tokamaks. In this brief report, preliminary results from the early stages of this effort are presented

Numerical determination of axisymmetric toroidal magnetohydrodynamic equilibria

International Nuclear Information System (INIS)

Johnson, J.L.; Dalhed, H.E.; Greene, J.M.

1978-07-01

Numerical schemes for the determination of stationary axisymmetric toroidal equilibria appropriate for modeling real experimental devices are given. Iterative schemes are used to solve the elliptic nonlinear partial differential equation for the poloidal flux function psi. The principal emphasis is on solving the free boundary (plasma-vacuum interface) equilibrium problem where external current-carrying toroidal coils support the plasma column, but fixed boundary (e.g., conducting shell) cases are also included. The toroidal current distribution is given by specifying the pressure and either the poloidal current or the safety factor profiles as functions of psi. Examples of the application of the codes to tokamak design at PPPL are given

Strong toroidal effects on tokamak tearing mode stability in the hybrid and conventional scenarios

International Nuclear Information System (INIS)

Ham, C J; Connor, J W; Cowley, S C; Gimblett, C G; Hastie, R J; Hender, T C; Martin, T J

2012-01-01

The hybrid scenario is thought to be an important mode of operation for the ITER tokamak. Analytic and numerical calculations demonstrate that toroidal effects at finite β have a strong influence on tearing mode stability of hybrid modes. Indeed, they persist in the large aspect ratio limit, R/a → ∞. A similar strong coupling effect is found between the m = 1, n = 1 harmonic and the m = 2, n = 1 harmonic if the minimum safety factor is less than unity. In both cases the tearing stability index, Δ′ increases rapidly as β approaches ideal marginal stability, providing a potential explanation for the onset of linearly unstable tearing modes. The numerical calculations have used an improved version of the T7 code (Fitzpatrick et al 1993 Nucl. Fusion 33 1533), and complete agreement is obtained with the analytic theory for this demanding test of the code. (paper)

Stability of Tokamaks with respect to slip motions

International Nuclear Information System (INIS)

Rebhan, E.; Salat, A.

1976-06-01

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.) [de

Compact tokamak reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

Theoretical study on the laser-driven ion-beam trace probe in toroidal devices with large poloidal magnetic field

Science.gov (United States)

Yang, X.; Xiao, C.; Chen, Y.; Xu, T.; Yu, Y.; Xu, M.; Wang, L.; Wang, X.; Lin, C.

2018-03-01

Recently, a new diagnostic method, Laser-driven Ion-beam Trace Probe (LITP), has been proposed to reconstruct 2D profiles of the poloidal magnetic field (Bp) and radial electric field (Er) in the tokamak devices. A linear assumption and test particle model were used in those reconstructions. In some toroidal devices such as the spherical tokamak and the Reversal Field Pinch (RFP), Bp is not small enough to meet the linear assumption. In those cases, the error of reconstruction increases quickly when Bp is larger than 10% of the toroidal magnetic field (Bt), and the previous test particle model may cause large error in the tomography process. Here a nonlinear reconstruction method is proposed for those cases. Preliminary numerical results show that LITP could be applied not only in tokamak devices, but also in other toroidal devices, such as the spherical tokamak, RFP, etc.

Influence of toroidal rotation on resistive tearing modes in tokamaks

International Nuclear Information System (INIS)

Wang, S.; Ma, Z. W.

2015-01-01

Influence of toroidal equilibrium plasma rotation on m/n = 2/1 resistive tearing modes is studied numerically using a 3D toroidal MHD code (CLT). It is found that the toroidal rotation with or without shear can suppress the tearing instability and the Coriolis effect in the toroidal geometry plays a dominant role on the rotation induced stabilization. For a high viscosity plasma (τ R /τ V  ≫ 1, where τ R and τ V represent resistive and viscous diffusion time, respectively), the effect of the rotation shear combined with the viscosity appears to be stabilizing. For a low viscosity plasmas (τ R /τ V  ≪ 1), the rotation shear shows a destabilizing effect when the rotation is large

Toroidal mode-conversion in the ICRF

International Nuclear Information System (INIS)

Jaun, A.; Hellsten, T.; Chiu, S.C.

1997-08-01

Mode-conversion is studied in the ion-cyclotron range of frequencies (ICRF) taking into account the toroidal geometry relevant for tokamaks. The global wavefields obtained using the gyrokinetic toroidal PENN code illustrate how the fast wave propagates to the neighborhood of the ion-ion hybrid resonance, where it is converted to a slow wave which deposits the wave energy through resonant interactions with the particles. The power deposition profiles obtained are dramatically different from the toroidal resonance absorption, showing that Budden's model is not a good approximation in the torus. Radially and poloidally localized wavefield structures characteristic of slow wave eigenmodes are predicted and could in experiments be driven to large amplitudes so as to interact efficiently with fast particles. (author) 5 figs., 1 tab., 48 refs

Influence of toroidal rotation on resistive tearing modes in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Wang, S.; Ma, Z. W., E-mail: zwma@zju.edu.cn [Institute for Fusion Theory and Simulation, Zhejiang University, Hangzhou 310027 (China)

2015-12-15

Influence of toroidal equilibrium plasma rotation on m/n = 2/1 resistive tearing modes is studied numerically using a 3D toroidal MHD code (CLT). It is found that the toroidal rotation with or without shear can suppress the tearing instability and the Coriolis effect in the toroidal geometry plays a dominant role on the rotation induced stabilization. For a high viscosity plasma (τ{sub R}/τ{sub V} ≫ 1, where τ{sub R} and τ{sub V} represent resistive and viscous diffusion time, respectively), the effect of the rotation shear combined with the viscosity appears to be stabilizing. For a low viscosity plasmas (τ{sub R}/τ{sub V} ≪ 1), the rotation shear shows a destabilizing effect when the rotation is large.

MHD-model for low-frequency waves in a tokamak with toroidal plasma rotation and problem of existence of global geodesic acoustic modes

Energy Technology Data Exchange (ETDEWEB)

Lakhin, V. P.; Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com, E-mail: vilkiae@gmail.com; Ilgisonis, V. I. [National Research Centre Kurchatov Institute (Russian Federation); Konovaltseva, L. V. [Peoples’ Friendship University of Russia (Russian Federation)

2015-12-15

A set of reduced linear equations for the description of low-frequency perturbations in toroidally rotating plasma in axisymmetric tokamak is derived in the framework of ideal magnetohydrodynamics. The model suitable for the study of global geodesic acoustic modes (GGAMs) is designed. An example of the use of the developed model for derivation of the integral conditions for GGAM existence and of the corresponding dispersion relation is presented. The paper is dedicated to the memory of academician V.D. Shafranov.

Helicity content and tokamak applications of helicity

International Nuclear Information System (INIS)

Boozer, A.H.

1986-05-01

Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities

Effects of 3D magnetic perturbations on toroidal plasmas

International Nuclear Information System (INIS)

Callen, J.D.

2011-01-01

Small three-dimensional (3D) magnetic field perturbations have many interesting and possibly useful effects on tokamak and quasi-symmetric stellarator plasmas. Plasma transport equations that include these effects, most notably on diamagnetic-level toroidal plasma flows, have recently been developed. The 3D field perturbations and their plasma effects can be classified according to their toroidal mode number n: low n (say 1-5) resonant (with field line pitch, q = m/n) and non-resonant fields, medium n (∼20, due to toroidal field ripple) and high n (due to microturbulence). Low n non-resonant fields induce a neoclassical toroidal viscosity (NTV) that damps toroidal rotation throughout the plasma towards an offset rotation in the counter-current direction. Recent tokamak experiments have generally confirmed and exploited these predictions by applying external low n non-resonant magnetic perturbations. Medium n toroidal field ripple produces similar effects plus possible ripple-trapping NTV effects and ion direct losses in the edge. A low n (e.g. n = 1) resonant field is mostly shielded by the toroidally rotating plasma at and inside the resonant (rational) surface. If it is large enough it can stop plasma rotation at the rational surface, facilitate magnetic reconnection there and lead to a growing stationary magnetic island (locked mode), which often causes a plasma disruption. Externally applied 3D magnetic perturbations usually have many components. In the plasma their lowest n (e.g. n = 1) externally resonant components can be amplified by kink-type plasma responses, particularly at high β. Low n plasma instabilities (e.g. resistive wall modes, neoclassical tearing modes) cause additional 3D magnetic perturbations in tokamak plasmas. Tearing modes in their nonlinear (Rutherford) regime bifurcate the topology and form magnetic islands. Finally, multiple resonant magnetic perturbations (RMPs) can, if not shielded by plasma rotation effects, cause local magnetic

Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

NARCIS (Netherlands)

J.W. Haverkort (Willem)

2013-01-01

htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma

Combined confinement system applied to tokamaks

International Nuclear Information System (INIS)

Ohkawa, Tihiro

1986-01-01

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

Tokamak simulation code manual

International Nuclear Information System (INIS)

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

1995-01-01

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

Numerical Tokamak Project code comparison

International Nuclear Information System (INIS)

Waltz, R.E.; Cohen, B.I.; Beer, M.A.

1994-01-01

The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

Computation of tokamak equilibria with steady flow

International Nuclear Information System (INIS)

Kerner, W.; Tokuda, Shinji

1987-08-01

The equations for ideal MHD equilibria with stationary flow are reexamined and addressed as numerically applied to tokamak configurations with a free plasma boundary. Both the isothermal (purely toroidal flow) and the poloidal flow cases are treated. Experiment-relevant states with steady flow (so far only in the toroidal direction) are computed by the modified SELENE40 code. (author)

Evidence for reduction of the toroidal ITG instability in the transition from saturated to improved Ohmic confinement in the tokamak TEXTOR

International Nuclear Information System (INIS)

Kreter, A; Schweer, B; Tokar, M Z; Unterberg, B

2003-01-01

In high density Ohmically heated discharges in the tokamak TEXTOR a transition from the saturated Ohmic confinement (SOC) to the improved Ohmic confinement (IOC) was observed triggered by a sudden reduction of the external gas flow. The SOC-IOC transition was investigated regarding the influence of the toroidal ITG instability driven by the ion temperature gradient (ITG). The ion temperature profiles were measured with high radial resolution by means of charge-exchange recombination spectroscopy (CXRS) with a high-energetic diagnostic hydrogen beam recently installed at TEXTOR. On the basis of the measured ion temperature distributions the η i parameter (ratio of the density and ion temperature decay lengths) and the growth rate of the toroidal ITG instability were calculated. After the SOC-IOC transition η i drops and lies in a noticeably smaller radial region over the threshold for the toroidal ITG. In consequence of it, the IOC regime is characterized by a clear reduction of the ITG growth rate γ ITG which was calculated including finite Larmor radius effects. The steepening of the plasma density profile after the decrease of the external gas flow is the main reason for the reduction of the ITG growth rate and the subsequent confinement transition to the IOC regime

Eddy current calculations for the Tore Supra toroidal field magnet

International Nuclear Information System (INIS)

Blum, J.

1983-01-01

An outline is given of the calculation of the eddy currents in the magnetic structures of a Tokamak, which can be assimilated to thin conductors, so that the three-dimensional problem can be reduced mathematically to a two-dimensional one, the variables being two orthogonal coordinates of the considered surface. A finite element method has been used in order to treat the complicated geometry of the set of the 18 toroidal field coil casings and mechanical structures of Tore Supra. This eddy current code has been coupled with an axisymmetric equilibrium code in order to simulate typical phases of a Tokamak discharge (plasma current rise, additional heating, disruption, cleaning discharge) and the losses in the toroidal field magnet have thus been calculated. (author)

Physical mechanism determining the radial electric field and its radial structure in a toroidal plasma

International Nuclear Information System (INIS)

Ida, Katsumi; Miura, Yukitoshi; Itoh, Sanae

1994-10-01

Radial structures of plasma rotation and radial electric field are experimentally studied in tokamak, heliotron/torsatron and stellarator devices. The perpendicular and parallel viscosities are measured. The parallel viscosity, which is dominant in determining the toroidal velocity in heliotron/torsatron and stellarator devices, is found to be neoclassical. On the other hand, the perpendicular viscosity, which is dominant in dictating the toroidal rotation in tokamaks, is anomalous. Even without external momentum input, both a plasma rotation and a radial electric field exist in tokamaks and heliotrons/torsatrons. The observed profiles of the radial electric field do not agree with the theoretical prediction based on neoclassical transport. This is mainly due to the existence of anomalous perpendicular viscosity. The shear of the radial electric field improves particle and heat transport both in bulk and edge plasma regimes of tokamaks. (author) 95 refs

Effect of toroidal plasma flow and flow shear on global MHD modes

International Nuclear Information System (INIS)

Chu, M.S.; Greene, J.M.; Jensen, T.H.; Miller, R.L.; Bondeson, A.; Johnson, R.W.; Mauel, M.E.

1995-01-01

The effect of a subsonic toroidal flow on the linear magnetohydrodynamic stability of a tokamak plasma surrounded by an external resistive wall is studied. A complex non-self-adjoint eigenvalue problem for the stability of general kink and tearing modes is formulated, solved numerically, and applied to high β tokamaks. Results indicate that toroidal plasma flow, in conjunction with dissipation in the plasma, can open a window of stability for the position of the external wall. In this window, stable plasma beta values can significantly exceed those predicted by the Troyon scaling law with no wall. Computations utilizing experimental data indicate good agreement with observations

Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

International Nuclear Information System (INIS)

Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

2014-03-01

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

1D equation for toroidal momentum transport in a tokamak

International Nuclear Information System (INIS)

Rozhansky, V A; Senichenkov, I Yu

2010-01-01

A 1D equation for toroidal momentum transport is derived for a given set of turbulent transport coefficients. The averaging is performed taking account of the poloidal variation of the toroidal fluxes and is based on the ambipolar condition of the zero net radial current through the flux surface. It is demonstrated that taking account of the Pfirsch-Schlueter fluxes leads to a torque in the toroidal direction which is proportional to the gradient of the ion temperature. This effect is new and has not been discussed before. The boundary condition at the separatrix, which is based on the results of the 2D simulations of the edge plasma, is formulated.

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

International Nuclear Information System (INIS)

Georgiou, G.E.

1979-01-01

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

Stochasticity and the m = 1 mode in tokamaks

International Nuclear Information System (INIS)

Izzo, R.; Monticello, D.A.; Stodiek, W.; Park, W.

1986-05-01

It has recently been proposed that stochasticity resulting from toroidal coupling could lead to a saturation of the m = 1 internal mode in tokamaks. We present results from the nonlinear evolution of the m = 1 mode with toroidal coupling that show that stochasticity is not enough to cause saturation of the m = 1 mode

Non-inductively driven tokamak plasmas at near-unity βt in the Pegasus toroidal experiment

Science.gov (United States)

Reusch, J. A.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Perry, J. M.; Pierren, C.; Rhodes, A. T.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Weberski, J. D.

2018-05-01

A major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓi, high elongation κ, and high toroidal and normalized beta ( βt and βN) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓi. The low aspect ratio ( R0/a ˜1.2 ) of Pegasus allows access to high κ and high normalized plasma currents ( IN=Ip/a BT>14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high βt plasmas. Equilibrium analyses indicate that βt up to ˜100% is achieved. These high βt discharges disrupt at the ideal no-wall β limit at βN˜7.

Toroidal Precession as a Geometric Phase

Energy Technology Data Exchange (ETDEWEB)

J.W. Burby and H. Qin

2012-09-26

Toroidal precession is commonly understood as the orbit-averaged toroidal drift of guiding centers in axisymmetric and quasisymmetric configurations. We give a new, more natural description of precession as a geometric phase effect. In particular, we show that the precession angle arises as the holonomy of a guiding center's poloidal trajectory relative to a principal connection. The fact that this description is physically appropriate is borne out with new, manifestly coordinate-independent expressions for the precession angle that apply to all types of orbits in tokamaks and quasisymmetric stellarators alike. We then describe how these expressions may be fruitfully employed in numerical calculations of precession.

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

International Nuclear Information System (INIS)

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

1990-05-01

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

Toroidal ripple transport of beam ions in the mega-ampère spherical tokamak

International Nuclear Information System (INIS)

McClements, K. G.; Hole, M. J.

2012-01-01

The transport of injected beam ions due to toroidal magnetic field ripple in the mega-ampère spherical tokamak (MAST) is quantified using a full orbit particle tracking code, with collisional slowing-down and pitch-angle scattering by electrons and bulk ions taken into account. It is shown that the level of ripple losses is generally rather low, although it depends sensitively on the major radius of the outer midplane plasma edge; for typical values of this parameter in MAST plasmas, the reduction in beam heating power due specifically to ripple transport is less than 1%, and the ripple contribution to beam ion diffusivity is of the order of 0.1 m 2 s –1 or less. It is concluded that ripple effects make only a small contribution to anomalous transport rates that have been invoked to account for measured neutron rates and plasma stored energies in some MAST discharges. Delayed (non-prompt) losses are shown to occur close to the outer midplane, suggesting that banana-drift diffusion is the most likely cause of the ripple-induced losses.

Influence of toroidal rotation on tearing modes

Science.gov (United States)

Cai, Huishan; Cao, Jintao; Li, Ding

2017-10-01

Tearing modes stability analysis including toroidal rotation is studied. It is found that rotation affects the stability of tearing modes mainly through the interaction with resistive inner region of tearing mode. The coupling of magnetic curvature with centrifugal force and Coriolis force provides a perturbed perpendicular current, and a return parallel current is induced to affect the stability of tearing modes. Toroidal rotation plays a stable role, which depends on the magnitude of Mach number and adiabatic index Γ, and is independent on the direction of toroidal rotation. For Γ >1, the scaling of growth rate is changed for typical Mach number in present tokamaks. For Γ = 1 , the scaling keeps unchanged, and the effect of toroidal rotation is much less significant, compared with that for Γ >1. National Magnetic Confinement Fusion Science Program and National Science Foundation of China under Grants No. 2014GB106004, No. 2013GB111000, No. 11375189, No. 11075161 and No. 11275260, and Youth Innovation Promotion Association CAS.

Boundary Plasma Turbulence Simulations for Tokamaks

International Nuclear Information System (INIS)

Xu, X.; Umansky, M.; Dudson, B.; Snyder, P.

2008-05-01

The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T e ; T i ) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics

Rotation profile flattening and toroidal flow shear reversal due to the coupling of magnetic islands in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Tobias, B.; Grierson, B. A.; Okabayashi, M. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Chen, M.; Domier, C. W.; Luhmann, N. C.; Muscatello, C. M. [University of California at Davis, Davis, California 95616 (United States); Classen, I. G. J. [Dutch Institute for Fundamental Fusion Energy Research, DIFFER, Rhinjuizen (Netherlands); Fitzpatrick, R. [University of Texas at Austin, Austin, Texas 78705 (United States); Olofsson, K. E. J.; Paz-Soldan, C. [General Atomics, San Diego, California 92121 (United States)

2016-05-15

The electromagnetic coupling of helical modes, even those having different toroidal mode numbers, modifies the distribution of toroidal angular momentum in tokamak discharges. This can have deleterious effects on other transport channels as well as on magnetohydrodynamic (MHD) stability and disruptivity. At low levels of externally injected momentum, the coupling of core-localized modes initiates a chain of events, whereby flattening of the core rotation profile inside successive rational surfaces leads to the onset of a large m/n = 2/1 tearing mode and locked-mode disruption. With increased torque from neutral beam injection, neoclassical tearing modes in the core may phase-lock to each other without locking to external fields or structures that are stationary in the laboratory frame. The dynamic processes observed in these cases are in general agreement with theory, and detailed diagnosis allows for momentum transport analysis to be performed, revealing a significant torque density that peaks near the 2/1 rational surface. However, as the coupled rational surfaces are brought closer together by reducing q{sub 95}, additional momentum transport in excess of that required to attain a phase-locked state is sometimes observed. Rather than maintaining zero differential rotation (as is predicted to be dynamically stable by single-fluid, resistive MHD theory), these discharges develop hollow toroidal plasma fluid rotation profiles with reversed plasma flow shear in the region between the m/n = 3/2 and 2/1 islands. The additional forces expressed in this state are not readily accounted for, and therefore, analysis of these data highlights the impact of mode coupling on torque balance and the challenges associated with predicting the rotation dynamics of a fusion reactor—a key issue for ITER.

The comparative analysis of the different mechanisms of toroidal rotation in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Sabot, R [Association Euratom-CEA, Centre d` Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Parail, V [Kurchatov Institute, Moscow (Russian Federation)

1994-07-01

The toroidal plasma rotation appears as one the possible mechanism for suppression of plasma turbulence. Several mechanisms are believed to contribute to the toroidal plasma rotation. The results of numerical simulation of the toroidal rotation on JET are presented, where are taken into consideration the following effects: the neoclassical viscosity due to banana and ripple trapped particles, the anomalous viscosity due to plasma turbulence, the momentum input by NBI (neutron beam injection) and ion momentum loss near the separatrix due to prompt ion losses. The NBI appeared to be the principal source of toroidal plasma rotation. 6 refs., 2 figs.

Integral torque balance in tokamaks

International Nuclear Information System (INIS)

Pustovitov, V.D.

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

Selcow, E.C.

1984-01-01

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

Computational studies of tokamak plasmas

International Nuclear Information System (INIS)

Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

1981-02-01

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

Linear wave propagation in a hot axisymmetric toroidal plasma

International Nuclear Information System (INIS)

Jaun, A.

1995-03-01

Kinetic effects on the propagation of the Alfven wave are studied for the first time in a toroidal plasma relevant for experiments. This requires the resolution of a set of coupled partial differential equations whose coefficients depend locally on the plasma parameters. For this purpose, a numerical wave propagation code called PENN has been developed using either a bilinear or a bicubic Hermite finite element discretization. It solves Maxwell's equations in toroidal geometry, with a dielectric tensor operator that takes into account the linear response of the plasma. Two different models have been implemented and can be used comparatively to describe the same physical case: the first treats the plasma as resistive fluids and gives results which are in good agreement with toroidal fluid codes. The second is a kinetic model and takes into account the finite size of the Larmor radii; it has successfully been tested against a kinetic plasma model in cylindrical geometry. New results have been obtained when studying kinetic effects in toroidal geometry. Two different conversion mechanisms to the kinetic Alfven wave have been described: one occurs at toroidally coupled resonant surfaces and is the kinetic counterpart of the fluid models' resonance absorption. The other has no such correspondence and results directly from the toroidal coupling between the kinetic Alfven wave and the global wavefield. An analysis of a heating scenario suggests that it might be difficult to heat a plasma with Alfven waves up to temperatures that are relevant for a tokamak reactor. Kinetic effects are studied for three types of global Alfven modes (GAE, TAE, BAE) and a new class of kinetic eigenmodes is described which appear inside the fluid gap: it could be related to recent observations in the JET (Joint European Torus) tokamak. (author) 56 figs., 6 tabs., 58 refs

Linear wave propagation in a hot axisymmetric toroidal plasma

Energy Technology Data Exchange (ETDEWEB)

Jaun, A [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP)

1995-03-01

Kinetic effects on the propagation of the Alfven wave are studied for the first time in a toroidal plasma relevant for experiments. This requires the resolution of a set of coupled partial differential equations whose coefficients depend locally on the plasma parameters. For this purpose, a numerical wave propagation code called PENN has been developed using either a bilinear or a bicubic Hermite finite element discretization. It solves Maxwell`s equations in toroidal geometry, with a dielectric tensor operator that takes into account the linear response of the plasma. Two different models have been implemented and can be used comparatively to describe the same physical case: the first treats the plasma as resistive fluids and gives results which are in good agreement with toroidal fluid codes. The second is a kinetic model and takes into account the finite size of the Larmor radii; it has successfully been tested against a kinetic plasma model in cylindrical geometry. New results have been obtained when studying kinetic effects in toroidal geometry. Two different conversion mechanisms to the kinetic Alfven wave have been described: one occurs at toroidally coupled resonant surfaces and is the kinetic counterpart of the fluid models` resonance absorption. The other has no such correspondence and results directly from the toroidal coupling between the kinetic Alfven wave and the global wavefield. An analysis of a heating scenario suggests that it might be difficult to heat a plasma with Alfven waves up to temperatures that are relevant for a tokamak reactor. Kinetic effects are studied for three types of global Alfven modes (GAE, TAE, BAE) and a new class of kinetic eigenmodes is described which appear inside the fluid gap: it could be related to recent observations in the JET (Joint European Torus) tokamak. (author) 56 figs., 6 tabs., 58 refs.

Structural analysis of TFTR toroidal field coil conceptual design

International Nuclear Information System (INIS)

Smith, R.A.

1975-10-01

The conceptual design evaluation of the V-shaped toroidal field coils on the Tokamak Fusion Test Reactor has been performed by detailed structural analysis with the finite element method. The innovation provided by this design and verified in this work is the capability to support toroidal field loads while simultaneously performing the function of twist restraint against the device axial torques resulting from the vertical field loads. The evaluations made for the conceptual design provide predictions for coil deflections and stresses. The results are available for the separate effects from toroidal fields, poloidal fields, and the thermal expansion of the coils as well as for the superposition of the primary loads and the primary plus thermal loads

Reynolds stress of localized toroidal modes

International Nuclear Information System (INIS)

Zhang, Y.Z.; Mahajan, S.M.

1995-02-01

An investigation of the 2D toroidal eigenmode problem reveals the possibility of a new consistent 2D structure, the dissipative BM-II mode. In contrast to the conventional ballooning mode, the new mode is poloidally localized at π/2 (or -π/2), and possesses significant radial asymmetry. The radial asymmetry, in turn, allows the dissipative BM-II to generate considerably larger Reynolds stress as compared to the standard slab drift type modes. It is also shown that a wide class of localized dissipative toroidal modes are likely to be of the dissipative BM-II nature, suggesting that at the tokamak edge, the fluctuation generated Reynolds stress (a possible source of poloidal flow) can be significant

Reynolds stress of localized toroidal modes

International Nuclear Information System (INIS)

Zhang, Y.Z.; Mahajan, S.M.

1995-01-01

An investigation of the 2D toroidal eigenmode problem reveals the possibility of a new consistent 2D structure, the dissipative BM-II mode. In contrast to the conventional ballooning mode, the new mode is poloidally localized at π/2 (or -π/2), and possesses significant radial asymmetry. The radial asymmetry, in turn, allows the dissipative BM-II to generate considerably larger Reynolds stress as compared to the standard slab drift type modes. It is also shown that a wide class of localized dissipative toroidal modes are likely to be of the dissipative BM-II nature, suggesting that at the tokamak edge, the fluctuation generated Reynolds stress (a possible source of poloidal flow) can be significant. (author). 15 refs

Flow shear stabilization of hybrid electron-ion drift mode in tokamaks

International Nuclear Information System (INIS)

Bai, L.

1999-01-01

In this paper, a model of sheared flow stabilization on hybrid electron-ion drift mode is proposed. At first, in the presence of dissipative trapped electrons, there exists an intrinsic oscillation mode in tokamak plasmas, namely hybrid dissipative trapped electron-ion temperature gradient mode (hereafter, called as hybrid electron-ion drift mode). This conclusion is in agreement with the observations in the simulated tokamak experiment on the CLM. Then, it is found that the coupling between the sheared flows and dissipative trapped electrons is proposed as the stabilization mechanism of both toroidal sheared flow and poloidal sheared flow on the hybrid electron-ion drift mode, that is, similar to the stabilizing effect of poloidal sheared flow on edge plasmas in tokamaks, in the presence of both dissipative trapped electrons and toroidal sheared flow, large toroidal sheared flow is always a strong stabilizing effect on the hybrid electron-ion drift mode in internal transport barrier location, too. This result is consistent with the experimental observations in JT-60U. (author)

Flow shear stabilization of hybrid electron-ion drift mode in tokamaks

International Nuclear Information System (INIS)

Bai, L.

2001-01-01

In this paper, a model of sheared flow stabilization on hybrid electron-ion drift mode is proposed. At first, in the presence of dissipative trapped electrons, there exists an intrinsic oscillation mode in tokamak plasmas, namely hybrid dissipative trapped electron-ion temperature gradient mode (hereafter, called as hybrid electron-ion drift mode). This conclusion is in agreement with the observations in the simulated tokamak experiment on the CLM. Then, it is found that the coupling between the sheared flows and dissipative trapped electrons is proposed as the stabilization mechanism of both toroidal sheared flow and poloidal sheared flow on the hybrid electron-ion drift mode, that is, similar to the stabilizing effect of poloidal sheared flow on edge plasmas in tokamaks, in the presence of both dissipative trapped electrons and toroidal sheared flow, large toroidal sheared flow is always a strong stabilizing effect on the hybrid electron-ion drift mode in internal transport barrier location, too. This result is consistent with the experimental observations in JT-60U. (author)

Lower hybrid heating experiments in tokamaks: an overview

International Nuclear Information System (INIS)

Porkolab, M.

1985-10-01

Lower hybrid wave propagation theory relevant to heating fusion grade plasmas (tokamaks) is reviewed. A brief discussion of accessibility, absorption, and toroidal ray propagation is given. The main part of the paper reviews recent results in heating experiments on tokamaks. Both electron and ion heating regimes will be discussed. The prospects of heating to high temperatures in reactor grade plasmas will be evaluated

ELECTRIC POTENIAL CELLS AT THE DIVERTED TOKAMAK SEPARATRIX

International Nuclear Information System (INIS)

SCHAFFER, M.J.; PORTER, G.D.; BOEDO, J.A.; BRAY, B.D.; HSIEH, C.L.; MOYER, R.A.; ROGNLIEN, T.D.; STANGEBY, P.C.; WATKINS, J.G.

2000-01-01

OAK-B135 Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p e ) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼ 50 V) drives E x B circulation (potential cell) of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arise from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the (del)B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. They speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode

Interaction of an ice pellet and a toroidal plasma in the JIPP T-IIU tokamak with the injection-angle controllable system

International Nuclear Information System (INIS)

Sato, K.N.; Sakakita, H.; Liang, R.; Hamada, Y.; Ida, K.; Kano, Y.; Sakamoto, M.

1994-01-01

The interaction of an ice pellet and a toroidal plasma has been studied in the JIPP T-IIU tokamak by using an injection-angle controllable system. In order to carry out various basic experiments by varying the pellet deposition profile within a plasma, anew technique for an ice pellet injection system with controllability of the injection angle has been developed and installed with the JIPP t-IIU tokamak. Injection angle can be varied easily and successfully during an interval of two plasma shots in the course of an experiment. The injection angle has been varied poloidally from 6 to 6 degree by changing the angle of the last stage drift tube, and this makes possible for pellets to aim at from about r = -2 a/3 to r = 2 a/3 of the plasma. From two dimensional observations by CCD cameras, details of the pellet ablation structures with various injections angles have been studied, and a couple of interesting phenomena have been found. In the case of an injection angle (θ) larger than a certain value (θ ≥ 4 0 ), a pellet penetrates straightly through the plasma with a trace of straight ablation cloud, which has been expected from usual theoretical consideration. On the other hand, a long helical tail of ablation light has been observed in the case of the angle smaller than the certain value (θ ≤ 4 0 ). The direction of helical rotation (tail) is independent to that of the total magnetic field lines of the torus. In order to examine the tail direction, further experiments have been carried out as to four conditions of the combination with two (clockwise and counter-clockwise) toroidal field directions and with two plasma current directions. The results show that it seems to rotate to the electron diamagnetic direction poloidally, and to the opposite to the plasma current direction toroidally. Consideration on various cross sections including charge exchange, ionization and elastic collisions leads us to the conclusion that the tail-shaped phenomena may come from

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Transport in the high temperature core of toroidal confinement systems

International Nuclear Information System (INIS)

Weiland, J.

1994-01-01

Recent theoretical and experimental results on confinement of hot plasmas in toroidal devices, particularly tokamaks, are discussed from general principal points of view and related to predictions from a toroidal drift wave model using a full transport matrix including off diagonal terms. A reactive fluid model corresponding to a two pole approximation of the kinetic response is used. This model has the ability to reproduce both adiabatic and isothermal limits of the perpendicular dynamics. 106 refs, 8 figs, 1 tab

Constrained ripple optimization of Tokamak bundle divertors

International Nuclear Information System (INIS)

Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

1983-02-01

Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ω B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple ( 0 ) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded

Modal analysis of a stiffened toroidal shell sector

International Nuclear Information System (INIS)

Cerreta, R.; Di Pietro, E.; Pizzuto, A.

1987-01-01

This paper presents the results of the modal analysis of a sector of the toroidal vacuum vessel of a new experimental machine for research in the field of controlled thermonuclear fusion (FTU - Frascati Tokamak Upgrade). The vacuum vessel, one of the most critical components of the experimental device, consist of 12 stainless steel toroidal sectors, and it is designed to withstand pulsed electromagnetic loads during operation. Results of the modal analysis of the stiffened toroidal shell sector are compared and discussed with regard to the experimental data. Theoretical eigenvalues and eigenvectors have been predicted by means of ABAQUS finite element code. Experimental analysis has been carried out on a full scale model and natural frequencies have been measured. Satisfactory agreement between experimental and theoretical eigenvalues has been found

Design, simulation and construction of the Taban tokamak

Science.gov (United States)

H, R. MIRZAEI; R, AMROLLAHI

2018-04-01

This paper describes the design and construction of the Taban tokamak, which is located in Amirkabir University of Technology, Tehran, Iran. The Taban tokamak was designed for plasma investigation. The design, simulation and construction of essential parts of the Taban tokamak such as the toroidal field (TF) system, ohmic heating (OH) system and equilibrium field system and their power supplies are presented. For the Taban tokamak, the toroidal magnetic coil was designed to produce a maximum field of 0.7 T at R = 0.45 m. The power supply of the TF was a 130 kJ, 0–10 kV capacitor bank. Ripples of toroidal magnetic field at the plasma edge and plasma center are 0.2% and 0.014%, respectively. For the OH system with 3 kA current, the stray field in the plasma region is less than 40 G over 80% of the plasma volume. The power supply of the OH system consists of two stages, as follows. The fast bank stage is a 120 μF, 0–5 kV capacitor that produces 2.5 kA in 400 μs and the slow bank stage is 93 mF, 600 V that can produce a maximum of 3 kA. The equilibrium system can produce uniform magnetic field at plasma volume. This system’s power supply, like the OH system, consists of two stages, so that the fast bank stage is 500 μF, 800 V and the slow bank stage is 110 mF, 200 V.

Experimental studies of compact toroids

International Nuclear Information System (INIS)

1991-01-01

The Berkeley Compact Toroid Experiment (BCTX) device is a plasma device with a Marshall-gun generated, low aspect ratio toroidal plasma. The device is capable of producing spheromak-type discharges and may, with some modification, produce low-aspect ratio tokamak configurations. A unique aspect of this experimenal devie is its large lower hybrid (LH) heating system, which consists of two 450MHz klystron tubes generating 20 megawatts each into a brambilla-type launching structure. Successful operation with one klystron at virtually full power (18 MW) has been accomplished with 110 μs pulse length. A second klystron is currently installed in its socket and magnet but has not been added to the RF drive system. This report describes current activities and accomplishments and describes the anticipated results of next year's activity

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

International Nuclear Information System (INIS)

1976-01-01

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

Summary report on tokamak confinement experiments

International Nuclear Information System (INIS)

1982-03-01

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

Configuration studies for a small-aspect-ratio tokamak stellarator hybrid

International Nuclear Information System (INIS)

Carreras, B.A.; Lynch, V.E.; Ware, A.

1996-08-01

The use of modulated toroidal coils offers a new path to the tokamak-stellarator hybrids. Low-aspect-ratio configurations can be found with robust vacuum flux surfaces and rotational transform close to the transform of a reverse-shear tokamak. These configurations have clear advantages in minimizing disruptions and their effect and in reducing tokamak current drive needs. They also allow the study of low-aspect-ratio effects on stellarator confinement in small devices

Tokamak m = 1 magnetohydrodynamic calculations in toroidal geometry using a full set of nonlinear resistive magnetohydrodynamic equations

International Nuclear Information System (INIS)

Charlton, L.A.; Carreras, B.A.; Holmes, J.A.; Lynch, V.E.

1988-01-01

The linear stability and nonlinear evolution of the resistive m = 1 mode in tokamaks is studied using a full set of resistive magnetohydrodynamic (MHD) equations in toroidal geometry. The modification of the linear and nonlinear properties of the mode by a combination of strong toroidal effects and low resistivity is the focus of this work. Linearly there is a transition from resistive kink to resistive tearing behavior as the aspect ratio and resistivity are reduced, and there is a corresponding modification of the nonlinear behavior, including a slowing of the island growth and development of a Rutherford regime, as the tearing regime is approached. In order to study the sensitivity of the stability and evolution to assumptions concerning the equation of state, two sets of full nonlinear resistive MHD equations (a pressure convection set and an incompressible set) are used. Both sets give more stable nonlinear behavior as the aspect ratio is reduced. The pressure convection set shows a transition from a Kadomtsev reconnection at large aspect ratio to a saturation at small aspect ratio. The incompressible set yields Kadomtsev reconnection for all aspect ratios, but with a significant lengthening of the reconnection time and development of a Rutherford regime at an aspect ratio approaching the transition from a resistive kink mode to a tearing mode. The pressure convection set gives an incomplete reconnection similar to that sometimes seen experimentally. The pressure convection set is, however, strictly justified only at high beta

Compact Commercial Tokamak Reactor (CCTR): a concept for a 500-MWe commercial-tokamak fusion system

International Nuclear Information System (INIS)

Gillen, T.J.

1980-11-01

A detailed set of self-consistent parameters and costs for the conceptual design of a Compact Commercial Tokamak Reactor (CCTR) is given. Several of the basic design features are the following: an ignited plasma with a major radius of 4.9 m and minor radius of 1.4 m; a net electrical output of 500 MW; a borated-water-cooled, stainless steel shield; and a toroidal field of 12 T at the coil. The design, which utilizes the Westinghouse computer code for the COsting And Sizing of D-T burning Tokamaks (COAST), mainly provides the sizes and geometries associated with the definition of the main component features for which a detailed engineering design can be effectively undertaken. Design study alternatives, including a neutral beam driven design option, a design option with a toroidal field of 13 T at the coil, and a tungsten-shielded option are considered for the CCTR. Also included is the conceptual design of a Compact Fusion Engineering Device

Design of the TPX outboard toroidal limiters

International Nuclear Information System (INIS)

Schaubel, K.M.; Anderson, P.M.; Baxi, C.B.

1995-01-01

The Tokamak Physics Experiment outboard limiter system incorporates the passive stabilizer plates, the ripple armor, the toroidal break and the support structures. These components are designed to withstand substantial steady state heat loads and high mechanical forces caused by plasma disruptions. The design of these components has been developed to deal with the challenging thermal, structural and remote handling requirements

Effects of toroidicity on resistive tearing modes

International Nuclear Information System (INIS)

Izzo, R.; Monticello, D.A.; Manickam, J.; Strauss, H.R.; Grimm, R.; McGuire, K.

1983-03-01

A reduced set of resistive MHD equations is solved numerically in three dimensions to study the stability of tokamak plasmas. Toroidal effects are included self-consistently to leading and next order in inverse aspect ratio, epsilon. The equations satisfy an energy integral. In addition, the momentum equation yields the Grad-Shafranov equation correct to all orders in epsilon. Low beta plasma are studied using several different q-profiles. In all cases, the linear growth rates are reduced by finite toroidicity. Excellent agreement with resistive PEST is obtianed. In some cases, toroidal effects lead to complete stabilization of the mode. Nonlinear results show smaller saturated island widths for finite aspect ratio compared to the cylindrical limit. If the current channel is wide enough so as to produce steep gradients towards the outside of the plasma, both the finite aspect ratio cases and cylindrical cases disrupt

TORFA - toroidal reactor for fusion applications

International Nuclear Information System (INIS)

Jassby, D.L.

1980-09-01

The near-term goal of the US controlled fusion program should be the development, for practical applications, of an intense, quasi-steady, reliable 14-MeV neutron source with an electrical utilization efficiency at least 10 times larger than the value characterizing beam/solid-target neutron generators. This report outlines a method for implementing that goal, based on tokamak fusion reactors featuring resistive toroidal-field coils designed for ease of demountability

Destabilization of a peeling-ballooning mode by a toroidal rotation in tokamaks

International Nuclear Information System (INIS)

Aiba, N.; Hirota, M.; Tokuda, S.; Furukawa, M.

2009-01-01

Full text: From the viewpoint of the heat load on the divertor, Type-I edge localized mode (ELM) needs to be suppressed or the amplitude of this ELM needs to be reduced. In JT-60U, some experimental results showed that the ELM frequency depends on the toroidal rotation, and the rapid rotation in the counter direction of the plasma current changes from Type-I ELM to Grassy ELM, whose frequency is high and the amplitude is small. Recent experimental and theoretical/numerical studies in a static system have identified that both Type-I and Grassy ELMs are considered ideal magnetohydrodynamic (MHD) modes destabilizing near the plasma surface, called peeling-ballooning modes. To investigate the mechanism of the change of ELM frequency by a toroidal rotation, theoretical and numerical analyses are important for understanding the toroidal rotation effects on the peeling-ballooning mode. Previous works about the toroidal rotation effect on the edge MHD stability have illustrated that the toroidal rotation with shear can destabilize low/intermediate-n (<50) modes but can stabilize high-n modes, where n is the toroidal mode number. The stabilization of the high-n mode can be understood qualitatively in analogy with the infinite-n ballooning mode case. However, the destabilizing mechanism of the low/intermediate-n mode is not still clarified, and to understand the stability property related to ELM suppression/mitigation, it is important to clarify this destabilizing mechanism. In this paper, we investigate numerically the destabilizing effect of a toroidal rotation on the peeling-ballooning mode with a newly developed code MINERVA, which solves the Frieman-Rotenberg equation. Particularly, we pay attention to the effect of the centrifuged force on not only equilibrium but also change of equation of motion. (author)

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

NARCIS (Netherlands)

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

2015-01-01

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

Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

2009-01-01

Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

Extension of the flow-rate-of-strain tensor formulation of plasma rotation theory to non-axisymmetric tokamaks

Energy Technology Data Exchange (ETDEWEB)

Stacey, W. M. [Georgia Institute of Technology, Atlanta, Georgia 30332 (United States); Bae, C. [National Fusion Research Institute, Daejoen (Korea, Republic of)

2015-06-15

A systematic formalism for the calculation of rotation in non-axisymmetric tokamaks with 3D magnetic fields is described. The Braginskii Ωτ-ordered viscous stress tensor formalism, generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry, and the resulting fluid moment equations provide a systematic formalism for the calculation of toroidal and poloidal rotation and radial ion flow in tokamaks in the presence of various non-axisymmetric “neoclassical toroidal viscosity” mechanisms. The relation among rotation velocities, radial ion particle flux, ion orbit loss, and radial electric field is discussed, and the possibility of controlling these quantities by producing externally controllable toroidal and/or poloidal currents in the edge plasma for this purpose is suggested for future investigation.

Mode structure and continuum damping of high-n toroidal Alfven eigenmodes

International Nuclear Information System (INIS)

Rosenbluth, M.N.; Berk, H.L.; Van Dam, J.W.; Lindberg, D.M.

1992-02-01

An asymptotic theory is described for calculating the mode structure and continuum damping of short wave-length toroidal Alfven eigenmodes (TAE). The formalism somewhat resembles the treatment used for describing low-frequency toroidal modes with singular structure at a rational surface, where an inner solution, which for the TAE mode has toroidal coupling, is matched to an outer toroidally uncoupled solution. A three-term recursion relation among coupled poloidal harmonic amplitudes is obtained, whose solution gives the structure of the global wavefunction and the complex eigenfrequency, including continuum damping. Both analytic and numerical solutions are presented. The magnitude of the damping is essential for determining the thresholds for instability driven by the spatial gradients of energetic particles (e.g., neutral beam-injected ions or fusion-product alpha particles) contained in a tokamak plasma

Long-wavelength microinstabilities in toroidal plasmas

International Nuclear Information System (INIS)

Tang, W.W.; Rewoldt, G.

1993-01-01

Realistic kinetic toroidal eigenmode calculations have been carried out to support a proper assessment of the influence of long-wavelength microturbulence on transport in tokamak plasmas. In order to efficiently evaluate large-scale kinetic behavior extending over many rational surfaces, significant improvements have been made to a toroidal finite element code used to analyze the fully two-dimensional (r,θ) mode structures of trapped-ion and toroidal ion temperature gradient (ITG) instabilities. It is found that even at very long wavelengths, these eigenmodes exhibit a strong ballooning character with the associated radial structure relatively insensitive to ion Landau damping at the rational surfaces. In contrast to the long-accepted picture that the radial extent of trapped-ion instabilities is characterized by the ion-gyroradius-scale associated with strong localization between adjacent rational surfaces, present results demonstrate that under realistic conditions, the actual scale is governed by the large-scale variations in the equilibrium gradients. Applications to recent measurements of fluctuation properties in TFTR L-mode plasmas indicate that the theoretical trends appear consistent with spectral characteristics as well as rough heuristic estimates of the transport level. Benchmarking calculations in support of the development of a three-dimensional toroidal gyrokinetic code indicate reasonable agreement with respect to both the properties of the eigenfunctions and the magnitude of the eigenvalues during the linear phase of the simulations of toroidal ITG instabilities

Ion temperature gradient modes in toroidal helical systems

Energy Technology Data Exchange (ETDEWEB)

Kuroda, T. [Graduate University for Advanced Studies, Toki, Gifu (Japan); Sugama, H.; Kanno, R.; Okamoto, M.

2000-04-01

Linear properties of ion temperature gradient (ITG) modes in helical systems are studied. The real frequency, growth rate, and eigenfunction are obtained for both stable and unstable cases by solving a kinetic integral equation with proper analytic continuation performed in the complex frequency plane. Based on the model magnetic configuration for toroidal helical systems like the Large Helical Device (LHD), dependences of the ITG mode properties on various plasma equilibrium parameters are investigated. Particularly, relative effects of {nabla}B-curvature drifts driven by the toroidicity and by the helical ripples are examined in order to compare the ITG modes in helical systems with those in tokamaks. (author)

Ion temperature gradient modes in toroidal helical systems

International Nuclear Information System (INIS)

Kuroda, T.; Sugama, H.; Kanno, R.; Okamoto, M.

2000-04-01

Linear properties of ion temperature gradient (ITG) modes in helical systems are studied. The real frequency, growth rate, and eigenfunction are obtained for both stable and unstable cases by solving a kinetic integral equation with proper analytic continuation performed in the complex frequency plane. Based on the model magnetic configuration for toroidal helical systems like the Large Helical Device (LHD), dependences of the ITG mode properties on various plasma equilibrium parameters are investigated. Particularly, relative effects of ∇B-curvature drifts driven by the toroidicity and by the helical ripples are examined in order to compare the ITG modes in helical systems with those in tokamaks. (author)

Global Hybrid Simulations of Energetic Particle-driven Modes in Toroidal Plasmas

International Nuclear Information System (INIS)

Fu, G.Y.; Breslau, J.; Fredrickson, E.; Park, W.; Strauss, H.R.

2004-01-01

Global hybrid simulations of energetic particle-driven MHD modes have been carried out for tokamaks and spherical tokamaks using the hybrid code M3D. The numerical results for the National Spherical Tokamak Experiments (NSTX) show that Toroidal Alfven Eigenmodes are excited by beam ions with their frequencies consistent with the experimental observations. Nonlinear simulations indicate that the n=2 mode frequency chirps down as the mode moves out radially. For ITER, it is shown that the alpha-particle effects are strongly stabilizing for internal kink mode when central safety factor q(0) is sufficiently close to unity. However, the elongation of ITER plasma shape reduces the stabilization significantly

Non-Inductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta in the Pegasus Toroidal Experiment

Science.gov (United States)

Reusch, Joshua

2017-10-01

A major goal of the spherical tokamak research program is accessing a state of low internal inductance li, high elongation κ, high toroidal and normalized beta (βt and βN) , and low collisionality without solenoidal current drive. A new local helicity injection (LHI) system in the lower divertor region of the ultra-low aspect ratio Pegasus ST provides non-solenoidally driven plasmas that exhibit most of these characteristics. LHI utilizes compact, edge-localized current sources (Ainj 4 cm2, Iinj 8 kA, Vinj 1.5 kV) for plasma startup and sustainment, and can sustain more than 200 kA of plasma current. Plasma growth via LHI is enhanced by a transition from a regime of high kink-like MHD activity to one of reduced MHD activity at higher frequencies and presumably shorter wavelengths. The strong edge current drive provided by LHI results in a hollow current density profile with low li. The low aspect ratio (R0 / a 1.2) of Pegasus allows ready access to high κ and MHD stable operation at very high normalized plasma currents (IN =Ip /aBT> 15). Thomson scattering measurements indicate Te 100 eV and ne 1 ×19 m-3. The impurity Ti evolution is correlated in time with high frequency magnetic fluctuations, implying substantial reconnection ion heating is driven by the applied helicity injection. Doppler spectroscopy indicates Ti >=Te and that the anomalous ion heating scales consistently with two fluid reconnection theory. Taken together, these features provide access to very high βt plasmas. Equilibrium analyses indicate βt up to 100% and βN 6.5 is achieved. At increasingly low BT, the discharge disrupts at the no-wall ideal stability limit. In these high βt discharges, a minimum |B| well forms over 50% of the plasma volume. This unique magnetic configuration may be of interest for testing predictions of stabilizing drift wave turbulence and/or improving energetic particle confinement. This work supported by US DOE Grants DE-FG02-96ER54375 and DE-SC0006928.

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

International Nuclear Information System (INIS)

Lee, V.D.

1987-01-01

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

Analysis on Θ pumping for tokamak current drive

International Nuclear Information System (INIS)

Miyamoto, Kenro; Naito, Osamu

1986-01-01

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

Energy storage for tokamak reactor cycles

International Nuclear Information System (INIS)

Buchanan, C.H.

1979-01-01

The inherent characteristic of a tokamak reactor requiring periodic plasma quench and reignition introduces the problem of energy storage to permit continuous electrical output to the power grid. The cycle under consideration in this paper is a 1000 second burn followed by a 100 second reignition phase. The physical size of a typical toroidal plasma reaction chamber for a tokamak reactor has been described earlier. The thermal energy storage requirements described in this reference will serve as a basis for much of the ensuing discussion

Absolute dissipative drift-wave instabilities in tokamaks

International Nuclear Information System (INIS)

Chen, L.; Chance, M.S.; Cheng, C.Z.

1979-07-01

Contrary to previous theoretical predictions, it is shown that the dissipative drift-wave instabilities are absolute in tokamak plasmas. The existence of unstable eigenmodes is shown to be associated with a new eigenmode branch induced by the finite toroidal couplings

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

International Nuclear Information System (INIS)

Chiyoda, Katsuji

1996-01-01

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

Stationary magnetohydrodynamic equilibrium of toroidal plasma in rotation

International Nuclear Information System (INIS)

Missiato, O.

1986-01-01

The stationary equations of classical magnetohydrodynamics are utilized to study the toroidal motion of a thermonuclear magnetically - confined plasma with toroidal symmetry (Tokamak). In the present work, we considered a purely toroidal stationary rotation and te problem is reduced to studing a second order partial differencial equation of eliptic type Maschke-Perrin. Assuming that the temperature remains constant on the magnetic surfaces, an analitic solution, valid for low Mach numbers (M ≤ 0 .4), was obtained for the above-mentioned equation by means of a technique developed by Pantuso Sudano. From the solution found, we traced graphs for the quantities which described the equilibrium state of the plasma, namely: mass density, pressure, temperature, electric current density and toroidal magnetic field. Finally we compare this analitical model with others works which utilized differents analitical models and numerical simulations. We conclude that the solutions obtained are in good agreement with the previos results. In addition, however, our model contains the results of Sudano-Goes with the additional advantage of employing much simple analitical expressions. (author) [pt

Influence of external toroidal flux on low-aspect-ratio toroidal plasma

International Nuclear Information System (INIS)

Ikuno, S.; Natori, M.; Kamitani, A.

1999-01-01

In the HIST device, the external flux is generated by two kinds of currents: the current I s flowing along the symmetry axis and the bias coil current I D . The influence of the external flux on the MHD equilibrium and stability of the low-aspect-ratio toroidal plasma in the HIST device is investigated numerically. Equilibrium configurations of the low-aspect-ratio toroidal plasma in the HIST device are numerically determined by means of the combination of FDM and BEM. The influence of I s and I D on their stability is also investigated by using the Mercier criterion. The results of computations show that the Mercier limit decreases to zero with increasing I s and with decreasing I D . Moreover, either a further increase in I s or a further decrease in I D raises the Mercier limit considerably. Besides, the equilibrium configuration in the HIST device changes its state from spheromak through ultra-low q to tokamak with increasing I s and with decreasing I D . (author)

3He functions in tokamak-pumped laser systems

International Nuclear Information System (INIS)

Jassby, D.L.

1986-10-01

3 He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the 3 He(n,p)T reaction, and thereby excite gaseous lasants mixed with the 3 He while simultaneously breeding tritium. The total 3 He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak

The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

Science.gov (United States)

Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

2017-10-01

Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

A steady-state axisymmetric toroidal system

International Nuclear Information System (INIS)

Hirano, K.

1984-01-01

Conditions for achieving a steady state in an axisymmetric toroidal system are studied with emphasis on a very-high-beta field-reversed configuration. The analysis is carried out for the electromotive force produced by the Ohkawa current that is induced by neutral-beam injection. It turns out that, since the perpendicular component of the current j-vectorsub(perpendicular) to the magnetic field can be generated automatically by the diamagnetic effect, only the parallel component j-vectorsub(parallel) must be driven by the electromotive force. The drive of j-vectorsub(parallel) generates shear in the field line so that the pure toroidal field on the magnetic axis is rotated towards the plasma boundary and matched to the external field lines. This matching condition determines the necessary amount of injection beam current and power. It is demonstrated that a very-high-beta field-reversed configuration requires only a small amount of current-driving beam power because almost all the toroidal current except that close to the magnetic axis is carried by the diamagnetic current due to high beta. A low-beta tokamak, on the other hand, needs very high current-driving power since most of the toroidal current is composed of j-vectorsub(parallel) which must be driven by the beam. (author)

Trapped ion mode in toroidally rotating plasmas

International Nuclear Information System (INIS)

Artun, M.; Tang, W.M.; Rewoldt, G.

1995-04-01

The influence of radially sheared toroidal flows on the Trapped Ion Mode (TIM) is investigated using a two-dimensional eigenmode code. These radially extended toroidal microinstabilities could significantly influence the interpretation of confinement scaling trends and associated fluctuation properties observed in recent tokamak experiments. In the present analysis, the electrostatic drift kinetic equation is obtained from the general nonlinear gyrokinetic equation in rotating plasmas. In the long perpendicular wavelength limit k τ ρ bi much-lt 1, where ρ bi is the average trapped-ion banana width, the resulting eigenmode equation becomes a coupled system of second order differential equations nmo for the poloidal harmonics. These equations are solved using finite element methods. Numerical results from the analysis of low and medium toroidal mode number instabilities are presented using representative TFTR L-mode input parameters. To illustrate the effects of mode coupling, a case is presented where the poloidal mode coupling is suppressed. The influence of toroidal rotation on a TFTR L-mode shot is also analyzed by including a beam species with considerable larger temperature. A discussion of the numerical results is presented

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.

Electronic system of TBR tokamak device

International Nuclear Information System (INIS)

Silva, R.P. da.

1980-01-01

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) [pt

Tokamak Engineering Technology Facility scoping study

International Nuclear Information System (INIS)

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

Advanced commercial tokamak study

International Nuclear Information System (INIS)

Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

1985-12-01

Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

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

International Nuclear Information System (INIS)

Conde, M.E.

1986-01-01

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

Hydrogen transport in a toroidal plasma using multigroup discrete-ordinates methodology

International Nuclear Information System (INIS)

Wienke, B.R.; Miller, W.F. Jr.; Seed, T.J.

1979-01-01

Neutral hydrogen transport in a fully ionized two-dimensional tokamak plasma was examined using discrete ordinates and contrasted with earlier analyses. In particular, curvature effects induced by toroidal geometries and ray effects caused by possible source localization were investigated. From an overview of the multigroup discrete-ordinates approximation, methodology in two-dimensional cylindrical geometry is detailed, mesh and plasma zoning procedures are sketched, and the piecewise polynomial solution algorithm on a triangular domain is obtained. Toroidal effects and comparisons as related to reaction rates and perticle spectra are examined for various model and source configurations

Field load and displacement boundary condition computer program used for the finite element analysis and design of toroidal field coils in a tokamak

International Nuclear Information System (INIS)

Smith, R.A.

1975-06-01

The design evaluation of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX) and the Tokamak Fusion Test Reactor (TFTR) has been performed by structural analysis with the finite element method. The technique employed has been simplified with supplementary computer programs that are used to generate the input data for the finite element computer program. Significant automation has been provided by computer codes in three areas of data input. These are the definition of coil geometry by a mesh of node points, the definition of finite elements via the node points and the definition of the node point force/displacement boundary conditions. The computer programs by name that have been used to perform the above functions are PDXNODE, ELEMENT and PDXFORC. The geometric finite element modeling options for toroidal field coils provided by PDXNODE include one-fourth or one-half symmetric sections of circular coils, oval shaped coils or dee-shaped coils with or without a beveled wedging surface. The program ELEMENT which defines the finite elements for input to the finite element computer code can provide considerable time and labor savings when defining the model of coils of non-uniform cross-section or when defining the model of coils whose material properties are different in the R and THETA directions due to the laminations of alternate epoxy and copper windings. The modeling features provided by the program ELEMENT have been used to analyze the PLT and the TFTR toroidal field coils with integral support structures. The computer program named PDXFORC is described. It computes the node point forces in a model of a toroidal field coil from the vector crossproduct of the coil current and the magnetic field. The model can be of one-half or one-fourth symmetry to be consistent with the node model defined by PDXNODE, and the magnetic field is computed from toroidal or poloidal coils

The magnet system of the Tokamak T-15 upgrade

International Nuclear Information System (INIS)

Khvostenko, P.P.; Azizov, E.A.; Alfimov, D.E.; Belyakov, V.A.; Bondarchuk, E.N.; Chudnovsky, A.N.; Dokuka, V.N.; Kavin, A.A.; Khayrutdinov, R.R.; Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N.; Lukash, V.E.; Mineev, A.B.; Muratov, V.P.

2015-01-01

Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.

The magnet system of the Tokamak T-15 upgrade

Energy Technology Data Exchange (ETDEWEB)

Khvostenko, P.P., E-mail: ppkhvost@rambler.ru [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Azizov, E.A.; Alfimov, D.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Belyakov, V.A.; Bondarchuk, E.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Chudnovsky, A.N.; Dokuka, V.N. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Kavin, A.A. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Khayrutdinov, R.R. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Lukash, V.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Mineev, A.B.; Muratov, V.P. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); and others

2015-10-15

Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.

2D full wave simulation on electromagnetic wave propagation in toroidal plasma

International Nuclear Information System (INIS)

Hojo, Hitoshi; Uruta, Go; Nakayama, Kazunori; Mase, Atsushi

2002-01-01

Global full-wave simulation on electromagnetic wave propagation in toroidal plasma with an external magnetic field imaging a tokamak configuration is performed in two dimensions. The temporal behavior of an electromagnetic wave launched into plasma from a wave-guiding region is obtained. (author)

Current drive by asymmetrical heating in a toroidal plasma

International Nuclear Information System (INIS)

Gahl, J.M.

1986-01-01

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

Toroidal coupling and frequency spectrum of tearing modes

International Nuclear Information System (INIS)

Edery, D.; Samain, A.

1989-05-01

The frequency spectrum of tearing modes is analyzed with the help of a mode coupling model including toroidal effects in the MHD regions and various non linear effects in the resonant layers. In particular it is shown that the sudden damping of the mode rotation and the simultaneous enhancement of the growth rate observed in tokamak, could be explained as a bifurcating solution of the dispersion equation

Tokamak Fusion Core Experiment (TFCX) special-purpose remote maintenance systems

International Nuclear Information System (INIS)

Masson, L.S.; Welland, H.J.

1985-01-01

A key element in the preconceptual design of the Tokamak Fusion Core Experiment (TFCX) was the development of design concepts for special-purpose remote maintenance systems. Included were systems for shield sector replacement, vacuum vessel sector and toroidal field coil replacement, limiter blade replacement, protective tile replacement, and general-purpose maintenance. This paper addresses these systems as they apply to the copper toroidal field (TF) coil version of the TFCX

Gyrokinetic global analysis of ion temperature gradient driven mode in reversed shear tokamaks

International Nuclear Information System (INIS)

Idomura, Y.; Tokuda, S.; Kishimoto, Y.

2003-01-01

A new toroidal gyrokinetic particle code has been developed to study the ion temperature gradient driven (ITG) turbulence in reactor relevant tokamak parameters. We use a new method based on a canonical Maxwellian distribution F CM (P φ , ε, μ), which is defined by three constants of motion in the axisymmetric toroidal system, the canonical angular momentum P φ , the energy ε, and the magnetic moment μ. A quasi-ballooning representation enables linear and nonlinear high-m,n global calculations with a good numerical convergence. Conservation properties are improved by using the optimized loading method. From comprehensive linear global analyses over a wide range of an unstable toroidal mode number spectrum (n=0∼100) in large tokamak parameters (a/ρ ti =320∼460), properties of the ITG modes in reversed shear tokamaks are discussed. In the nonlinear simulation, it is found that a new method based on F CM can simulate a zonal flow damping correctly, and spurious zonal flow oscillations, which are observed in a conventional method based on a local Maxwellian distribution F LM (ψ, ε, μ), do not appear in the nonlinear regime. (author)

Toroidal rotation braking with n = 1 magnetic perturbation field on JET

DEFF Research Database (Denmark)

Sun, Y; Liang, Y; Koslowski, H R

2010-01-01

A strong toroidal rotation braking has been observed in plasmas with application of an n = 1 magnetic perturbation field on the JET tokamak. Calculation results from the momentum transport analysis show that the torque induced by the n = 1 perturbation field has a global profile. The maximal value...

Tearing mode analysis in tokamaks, revisited

International Nuclear Information System (INIS)

Nishimura, Y.; Callen, J.D.; Hegna, C.C.

1997-12-01

A new Δ' shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio (ε ≤ 0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of finite pressure term. Numerical results compare favorably with Furth et al. results. The effects of finite pressure, which are shown to decrease Δ', are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric element stabilizes the tearing mode significantly, even in a low β regime before the toroidal magnetic curvature effects come into play. Double tearing modes in toroidal geometries are examined as well. Furthermore, m ≥ 2 tearing mode stability criteria are compared with three dimensional initial value MHD simulation by the FAR code

/sup 3/He functions in tokamak-pumped laser systems

Energy Technology Data Exchange (ETDEWEB)

Jassby, D.L.

1986-10-01

/sup 3/He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the /sup 3/He(n,p)T reaction, and thereby excite gaseous lasants mixed with the /sup 3/He while simultaneously breeding tritium. The total /sup 3/He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak.

Toroidal effects on the non-linearly saturated m = 1 island in tokamaks

International Nuclear Information System (INIS)

Avinash, K.; Haas, F.A.; Thyagaraja, A.

1990-01-01

This paper investigates the influence of toroidal effects (due to the coupling of various poloidal harmonics) on the non-linear saturation of the m=1 island. Bounds are obtained relating the aspect ratio, the shear at the q=1 surface and the saturated island width. Provided these bounds are satisfied, then we find that the cylindrical m=1 island theory is valid for toroidal geometry. (author)

Modular coils and finite-β operation of a quasi-axially symmetric tokamak

International Nuclear Information System (INIS)

Drevlak, M.

1998-01-01

Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nuehrenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (MERKEL, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (HIRSHMAN, S.P., VAN RIJ, W.I., MERKEL, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak

Feedback control of resistive wall modes in toroidal devices

International Nuclear Information System (INIS)

Liu, Y.Q.

2002-01-01

Active feedback of resistive wall modes is investigated using cylindrical theory and toroidal calculations. For tokamaks, good performance is obtained by using active coils with one set of coils in the poloidal direction and sensors detecting the poloidal field inside the first wall, located at the outboard mid-plane. With suitable width of the feedback coil such a system can give robust control with respect to variations in plasma current, pressure and rotation. Calculations are shown for ITER-like geometry with a double wall. The voltages and currents in the active coils are well within the design limits for ITER. Calculations for RFP's are presented for a finite number of coils both in the poloidal and toroidal directions. With 4 coils in the poloidal and 24 coils in the toroidal direction, all non-resonant modes can be stabilized both at high and low theta. Several types of sensors, including radial and internal poloidal or toroidal sensors, can stabilize the RWM, but poloidal sensors give the most robust performance. (author)

Aspects of the equilibrium and stability of counterstreaming-ion tokamaks

International Nuclear Information System (INIS)

Cordey, J.G.; Haas, F.A.

1976-01-01

An anisotropic high-β equilibrium is derived for the counterstreaming-beam tokamak (CBT). The critical β of the CBT is found to be of comparable magnitude to that occurring in a similar model of a scalar-pressure tokamak. It is shown that the toroidal current which is essential for equilibrium can be maintained by the counterstreaming ions. Finally, a brief discussion of the stability of the device is given. (author)

TFTR toroidal field coil design

International Nuclear Information System (INIS)

Smith, G.E.; Punchard, W.F.B.

1977-01-01

The design of the Tokamak Fusion Test Reactor (TFTR) Toroidal Field (TF) magnetic coils is described. The TF coil is a 44-turn, spiral-wound, two-pancake, water-cooled configuration which, at a coil current of 73.3 kiloamperes, produces a 5.2-Tesla field at a major radius of 2.48 meters. The magnetic coils are installed in titanium cases, which transmit the loads generated in the coils to the adjacent supporting structure. The TFTR utilizes 20 of these coils, positioned radially at 18 0 intervals, to provide the required toroidal field. Because it is very highly loaded and subject to tight volume constraints within the machine, the coil presents unique design problems. The TF coil requirements are summarized, the coil configuration is described, and the problems highlighted which have been encountered thus far in the coil design effort, together with the development tests which have been undertaken to verify the design

Advanced Toroidal Facility

International Nuclear Information System (INIS)

Johnson, R.L.

1985-01-01

The Advanced Toroidal Facility (ATF) is a new magnetic confinement plasma device under construction at the Oak Ridge National Laboratory (ORNL) that will lead to improvements in toroidal magnetic fusion reactors. The ATF is a type of stellerator, known as a ''torsatron'' which theoretically has the capability to operate at greater than or equal to8% beta in steady state. The ATF plasma has a major radius of 2.1 m, an average minor radius of 0.3 m, and a field of 2 T for a 2 s duration or 1 T steady state. The ATF device consists of a helical field (HF) coil set, a set of poloidal field (PF) coils, an exterior shell structure to support the coils, and a thin, helically contoured vacuum vessel inside the coils. The ATF replaces the Impurities Studies Experiment (ISX-B) tokamak at ORNL and will use the ISX-B auxiliary systems including 4 MW of electron cyclotron heating. The ATF is scheduled to start operation in late 1986. An overview of the ATF device is presented, including details of the construction process envisioned. 9 refs., 7 figs., 3 tabs

Technological start of T-15 tokamak. The start-up diagnostic complex

International Nuclear Information System (INIS)

Notkin, G.E.

1989-01-01

The T-15 tokamak with superconducting toroidal winding reached the technological start-up phase. The results of the first operating tests of the main tokamak components are reported. Due to improper function of both the vacuum and the cryogenic system, the nominal parameters of the vacuum and of the toroidal magnetic field have not been achieved. The non-optimum vacuum conditions made the discharge start-up difficult even when a pre-ionizing electron beam and a gyrotron generator were used. The pre-discharge plasma parametes were studied by means of a limited set of plasma diagnostic apparatus. Due to substantially deteriorated vacuum conditions, it was not possible to repeat the only one successful discharge with a current of 100 kA, lasting for 50 ms. (J.U.)

Turbulence induced radial transport of toroidal momentum in boundary plasma of EAST tokamak

International Nuclear Information System (INIS)

Zhao, N.; Yan, N.; Xu, G. S.; Wang, H. Q.; Wang, L.; Ding, S. Y.; Chen, R.; Chen, L.; Zhang, W.; Hu, G. H.; Shao, L. M.; Wang, Z. X.

2016-01-01

Turbulence induced toroidal momentum transport in boundary plasma is investigated in H-mode discharge using Langmuir-Mach probes on EAST. The Reynolds stress is found to drive an inward toroidal momentum transport, while the outflow of particles convects the toroidal momentum outwards in the edge plasma. The Reynolds stress driven momentum transport dominates over the passive momentum transport carried by particle flux, which potentially provides a momentum source for the edge plasma. The outflow of particles delivers a momentum flux into the scrape-off layer (SOL) region, contributing as a momentum source for the SOL flows. At the L-H transitions, the outward momentum transport suddenly decreases due to the suppression of edge turbulence and associated particle transport. The SOL flows start to decelerate as plasma entering into H-mode. The contributions from turbulent Reynolds stress and particle transport for the toroidal momentum transport are identified. These results shed lights on the understanding of edge plasma accelerating at L-H transitions.

Exhaust, ELM and Halo physics using the MAST tokamak

International Nuclear Information System (INIS)

Counsell, G.F.; Ahn, J-W.; Kirk, A.; Helander, P.; Martin, R.; Tabasso, A.; Wilson, H.R.; Cohen, R.H.; Ryutov, D.D.; Yang, Y.

2003-01-01

The scrape-off layer (Sol) and divertor target plasma of a large spherical tokamak (ST) is characterised in detail for the first time. Scalings for the SOL heat flux width in MAST are developed and compared to conventional tokamaks. Modelling reveals the significance of the mirror force to the ST SOL. Core energy losses, including during ELMs, in MAST arrive predominantly (>80%) to the outboard targets in all geometries. Convective transport dominates energy losses during ELMs and MHD analysis suggests ELMs in MAST are Type III even at auxiliary heating powers well above the L-H threshold. ELMs are associated with a poloidally and/or toroidally localised radial efflux at ∼1 km/s well into the far SOL but not with any broadening of the target heat flux width. Toroidally asymmetric divertor biasing experiments have been conducted in an attempt to broaden the target heat flux width, with promising results. During vertical displacement events, the maximum product of the toroidal peaking factor and halo current fraction remains below 0.3, around half the ITER design limit. Evidence is presented that the resistance of the halo current path may have an impact on the distribution of halo current. (author)

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Ripple induced trapped particle loss in tokamaks

International Nuclear Information System (INIS)

White, R.B.

1996-05-01

The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks

Nonlinear equilibrium in Tokamaks including convective terms and viscosity

International Nuclear Information System (INIS)

Martin, P.; Castro, E.; Puerta, J.

2003-01-01

MHD equilibrium in tokamaks becomes very complex, when the non-linear convective term and viscosity are included in the momentum equation. In order to simplify the analysis, each new term has been separated in type gradient terms and vorticity depending terms. For the special case in which the vorticity vanishes, an extended Grad-Shafranov type equation can be obtained. However now the magnetic surface is not isobars or current surfaces as in the usual Grad-Shafranov treatment. The non-linear convective terms introduces gradient of Bernoulli type kinetic terms . Montgomery and other authors have shown the importance of the viscosity terms in tokamaks [1,2], here the treatment is carried out for the equilibrium condition, including generalized tokamaks coordinates recently described [3], which simplify the equilibrium analysis. Calculation of the new isobar surfaces is difficult and some computation have been carried out elsewhere for some particular cases [3]. Here, our analysis is extended discussing how the toroidal current density, plasma pressure and toroidal field are modified across the midplane because of the new terms (convective and viscous). New calculations and computations are also presented. (Author)

Feedback control for magnetic island suppression in tokamaks

NARCIS (Netherlands)

Hennen, B.A.

2011-01-01

A real-time feedback control system has been developed that finds, tracks, suppresses and/or stabilizes resistive magnetic instabilities in a nuclear fusion plasma. In a tokamak, magnetic fields confine a fusion plasma in a topology of toroidally nested magnetic surfaces. The power produced by the

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Neoclassical dissipation and resistive wall modes in tokamaks

International Nuclear Information System (INIS)

Shaing, K.C.

2004-01-01

It is shown that the critical toroidal plasma flow speed that is required to stabilize the resistive wall mode in tokamaks is reduced by a factor of the order of B/B θ or of 1.265ε 3sol4 B/B θ depending on the plasma parameters when the perturbed neoclassical viscosity driven current is taken into account. Here, B is the magnetic field strength, B θ is the poloidal magnetic field strength, and ε is the inverse aspect ratio. This effect is illustrated using an existing model for the resistive wall modes by including the neoclassical dissipation in the derivation of the dispersion relation. The derivation is based on fluid equations with the plasma viscosity, calculated using kinetic equation, as the closure. The reduction of the critical toroidal speed is a consequence of the parallel (to the magnetic field B) momentum equation when neoclassical viscosity becomes important. The results are compared with experimental observations in tokamaks

Tore-Supra: a Tokamak with superconducting toroidal field coils

International Nuclear Information System (INIS)

Turck, B.

1987-07-01

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

On the simulation of the tokamak longitUdinal field

International Nuclear Information System (INIS)

Simakov, A.S.

1978-01-01

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

Edge kink/ballooning mode stability in tokamaks with separatrix

International Nuclear Information System (INIS)

Medvedev, S Yu; Martynov, A A; Martin, Y R; Sauter, O; Villard, L

2006-01-01

Stability limits against external kink modes driven by large current density and pressure gradient values in the pedestal region are investigated for tokamak plasmas with separatrix. Stability diagrams for modes with different toroidal wave numbers under variations of pressure gradient and current density in the pedestal region are presented for several equilibrium configurations related to TCV. A scaling for the toroidal wave number of the most unstable mode is proposed. The influence of the plasma cross-section geometry on the stability limits is discussed

An experimental investigation of the propagation of a compact toroid along curved drift tubes

International Nuclear Information System (INIS)

Fukumoto, N.; Inoo, Y.; Nomura, M.; Nagata, M.; Uyama, T.; Ogawa, H.; Kimura, H.; Uehara, U.; Shibata, T.; Kashiwa, Y.; Suzuki, S.; Kasai, S.

2004-01-01

Compact toroid (CT) injection is a viable technology for fuelling large tokamak reactors in the future. Experimental demonstration of CT injection has thus far been conducted using horizontal injection in the midplane of tokamak devices. However, recent analyses indicate adverse effects of the toroidal magnetic field on CT injection. In order to avoid these adverse effects, the CT would need to be injectable in any direction. We have therefore devised a curved drift tube to change the direction of CT propagation and have experimentally demonstrated its efficacy. It has been observed that a CT can be transported smoothly through curved drift tubes with 45 deg. and 90 deg. bends without any appreciable change in the CT parameters. The magnetic field, electron density and speed of CTs transported through both 45 deg. and 90 deg. bends are similar to those observed in a linear drift tube. (author)

Global stability of plasmas with helical boundary deformation and net toroidal current against n=1,2 external modes

International Nuclear Information System (INIS)

Ardela, A.; Cooper, W.A.

1996-01-01

In this paper we resume a numerical study of the global stability of plasma with helical boundary deformation and non null net toroidal current. The aim was to see whether external modes with n=1,2 (n toroidal mode number) can be stabilized at values of β inaccessible to the tokamak. L=2,3 configurations with several aspect ratios and different numbers of equilibrium field periods are considered. A large variety of toroidal current densities and different pressure profiles are taken into account. Mercier stability is also investigated. (author) 4 figs., 6 refs

Resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1980-11-01

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

Design of a microwave calorimeter for the microwave tokamak experiment

International Nuclear Information System (INIS)

Marinak, M.

1988-01-01

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

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

International Nuclear Information System (INIS)

Azizov, E A

2012-01-01

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. (conferences and symposia)

Toroidal modeling of plasma response and resonant magnetic perturbation field penetration

Czech Academy of Sciences Publication Activity Database

Liu, Y.Q.; Kirk, A.; Sun, Y.; Cahyna, Pavel; Chapman, I.T.; Denner, P.; Fishpool, G.; Garofalo, A.M.; Harrison, J.R.; Nardon, E.

2012-01-01

Roč. 54, č. 12 (2012), s. 124013-124013 ISSN 0741-3335 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * resonant magnetic perturbation * neoclassical toroidal viscosity Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.369, year: 2012 http://iopscience.iop.org/0741-3335/54/12/124013/pdf/0741-3335_54_12_124013.pdf

Tokamak power systems studies at ANL

International Nuclear Information System (INIS)

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

1986-01-01

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

Tokamak Physics Experiment (TPX) power supply design and development

International Nuclear Information System (INIS)

Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.

1995-01-01

The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes

Tokamak rotation and charge exchange

International Nuclear Information System (INIS)

Hazeltine, R.D.; Rowan, W.L.; Solano, E.R.; Valanju, P.M.

1991-01-01

In the absence of momentum input, tokamak toroidal rotation rates are typically small - no larger in particular than poloidal rotation - even when the radial electric field is strong, as near the plasma edge. This circumstance, contradicting conventional neoclassical theory, is commonly attributed to the rotation damping effect of charge exchange, although a detailed comparison between charge-exchange damping theory and experiment is apparently unavailable. Such a comparison is attempted here in the context of recent TEXT experiments, which compare rotation rates, both poloidal and toroidal, in helium and hydrogen discharges. The helium discharges provide useful data because they are nearly free of ion-neutral charge exchange; they have been found to rotate toroidally in reasonable agreement with neoclassical predictions. The hydrogen experiments show much smaller toroidal motion as usual. The theoretical calculation uses the full charge-exchange operator and assumes plateau collisionality, roughly consistent with the experimental conditions. The authors calculate the ion flow as a function of v cx /v c , where v cx is the charge exchange rate and v c the Coulomb collision frequency. The results are in reasonable accord with the observations. 1 ref

Complete suppression of Pfirsch-Schlueter current in a toroidal l=3 stellarator

International Nuclear Information System (INIS)

Sato, Yasuhiko; Wakatani, Masahiro; Yokoyama, Masayuki; Pustovitov, V.D.

1999-10-01

Pfirsch-Schlueter (P-S) current is an inherent property of a finite pressure toroidal equilibrium of tokamak and stellarator. However, it was pointed out recently (V.D. Pustovitov, Nuclear Fusion 36 (1996) 583) that the P-S current would be suppressed completely if the external vertical field could be adjusted to satisfy the condition Ω= in an l=3 stellarator. Here Ω= 2 >/B 0 2 -2ε cosθ, l is a pole number, |B tilde| the vacuum helical magnetic field, B 0 the toroidal field, ε the inverse aspect ratio, θ the poloidal angle and denotes the average over the toroidal angle. An example of such a stellarator equilibrium is presented in this paper. For this stellarator equilibrium, behavior of rotational transform and Boozer magnetic spectrum is clarified when the pressure is increased. Both formation of helical magnetic axis and reduction of toroidal curvature are important ingredients to reduce the P-S current. However, the collisionless particle confinement is not improved in this example. (author)

Numerical stress analysis of toroidal coil by three-dimensional finite element method

International Nuclear Information System (INIS)

Nishimura, Hidetomo; Shimamoto, Susumu

1977-10-01

A structure analysis program based on finite element method for toroidal coils, developed in JAERI, and its example application to a medium-size tokamak are described. In this application, the effects of material anisotropy, poloidal field and spring constant value were studied, and also the influence of toroidal coil failure on the peak stress. The following were revealed. The effect of anisotropy on the peak stress in reinforcement must be considered. The effect of poloidal field on the peak stress is small compared with that of toroidal field. The spring constant value between coil and support does not much influence the peak stress value, The peak stress in reinforcement rises with increasing number of failed coils. In the case of 2000 nodes on the structure, CPU time with the program is about 40 min. (auth.)

Plasma rotation under a driven radial current in a tokamak

International Nuclear Information System (INIS)

Chang, C.S.

1999-01-01

The neoclassical behaviour of plasma rotation under a driven radial electrical current is studied in a tokamak geometry. An ambipolar radial electric field develops instantly in such a way that the driven current is balanced by a return current j p in the plasma. The j p x B torque pushes the plasma into a new rotation state both toroidally and poloidally. An anomalous toroidal viscosity is needed to avoid an extreme toroidal rotation speed. It is shown that the poloidal rotation relaxes to a new equilibrium speed, which is in general smaller than the E x B poloidal speed, and that the timescale for the relaxation of poloidal rotation is the same as that of toroidal rotation generation, which is usually much longer than the ion-ion collision time. (author)

Tokamak reactor for treating fertile material or waste nuclear by-products

Science.gov (United States)

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Magnet design considerations for Tokamak fusion reactors

International Nuclear Information System (INIS)

Purcell, J.R.; Chen, W.; Thomas, R.

1976-01-01

Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

Fast wave current drive experiment on the DIII-D tokamak

International Nuclear Information System (INIS)

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

1992-06-01

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

Calculation of modification to the toroidal magnetic field of the Tokamak Novillo. Part II

International Nuclear Information System (INIS)

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

1992-03-01

In a cylindrical magnetic topology. the confined plasma experiences 'classic' collisional transport phenomena. When bending the cylinder with the purpose of forming a toro, the magnetic field that before was uniform now it has a radial gradient which produces an unbalance in the magnetic pressure that is exercised on the plasma in the transverse section of the toro. This gives place to transport phenomena call 'neo-classicist'. In this work the structure of the toroidal magnetic field produced by toroidal coils of triangular form, to which are added even of coils of compensation with form of half moon is analyzed. With this type of coils it is looked for to minimize the radial gradient of the toroidal magnetic field. The values and characteristics of B (magnetic field) in perpendicular planes to the toro in different angular positions in the toroidal direction, looking for to cover all the cases of importance are exhibited. (Author)

Simulation experiment on magnetic field reconnection processes in tokamak

International Nuclear Information System (INIS)

Kiwamoto, Y.

1982-01-01

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

Advanced Toroidal Facility (ATF)

International Nuclear Information System (INIS)

Thompson, P.B.

1985-01-01

The Advanced Toroidal Facility (ATF) is a new magnetic plasma confinement device, under construction at Oak Ridge National Laboratory (ORNL), which will lead to improvements in toroidal magnetic fusion reactors. ATF is a type of stellarator known as a torsatron which theoretically has the capability at greater than or equal to8% beta in steady state. The ATF plasma has a major radius of 2.1 m, an average minor radius of 0.3 m, and a field of 2 T for a 5-s duration or 1 T steady state. The ATF device consists of a helical field (HF) coil set, a set of poloidal field (PF) coils, an exterior shell structure to support the coils, and a thin helically contoured vacuum vessel inside the coils. The ATF replaces the ISX-B tokamak at ORNL and will use the ISX-B auxiliary systems including 4 MW of neutral injection heating and 0.2 MW of electron cyclotron heating. ATF device is scheduled to start operation in the fall of 1986. An overview of the ATF device is presented including details of the construction process envisioned

Closed expressions for the magnetic field of toroidal multipole configurations

International Nuclear Information System (INIS)

Sheffield, G.V.

1983-04-01

Closed analytic expressions for the vector potential and the magnetic field for the lower order toroidal multipoles are presented. These expressions can be applied in the study of tokamak plasma cross section shaping. An example of such an application is included. These expressions also allow the vacuum fields required for plasma equilibrium to be specified in a general form independent of a particular coil configuration

MHD equilibrium with toroidal rotation

International Nuclear Information System (INIS)

Li, J.

1987-03-01

The present work attempts to formulate the equilibrium of axisymmetric plasma with purely toroidal flow within ideal MHD theory. In general, the inertial term Rho(v.Del)v caused by plasma flow is so complicated that the equilibrium equation is completely different from the Grad-Shafranov equation. However, in the case of purely toroidal flow the equilibrium equation can be simplified so that it resembles the Grad-Shafranov equation. Generally one arbitrary two-variable functions and two arbitrary single variable functions, instead of only four single-variable functions, are allowed in the new equilibrium equations. Also, the boundary conditions of the rotating (with purely toroidal fluid flow, static - without any fluid flow) equilibrium are the same as those of the static equilibrium. So numerically one can calculate the rotating equilibrium as a static equilibrium. (author)

Effects of q and high beta on tokamak stability

International Nuclear Information System (INIS)

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

1984-08-01

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

Design of charge exchange recombination spectroscopy for the joint Texas experimental tokamak

Energy Technology Data Exchange (ETDEWEB)

Chi, Y.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Cheng, Z. F.; Hou, S. Y.; Cheng, C.; Li, Z.; Wang, J. R.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-11-15

The old diagnostic neutral beam injector first operated at the University of Texas at Austin is ready for rejoining the joint Texas experimental tokamak (J-TEXT). A new set of high voltage power supplies has been equipped and there is no limitation for beam modulation or beam pulse duration henceforth. Based on the spectra of fully striped impurity ions induced by the diagnostic beam the design work for toroidal charge exchange recombination spectroscopy (CXRS) system is presented. The 529 nm carbon VI (n = 8 − 7 transition) line seems to be the best choice for ion temperature and plasma rotation measurements and the considered hardware is listed. The design work of the toroidal CXRS system is guided by essential simulation of expected spectral results under the J-TEXT tokamak operation conditions.

Toroidal charge exchange recombination spectroscopy on EAST

Energy Technology Data Exchange (ETDEWEB)

Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Yingying [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Yu, Yi [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230026 (China); Shi, Yuejiang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230026 (China); WCI for Fusion Theory, National Fusion Research Institute, 52 Eoeun-Dong, Yusung-Gu, Daejeon 305-333 (Korea, Republic of); Lyu, Bo; Fu, Jia [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Du, Xuewei; Yin, Xianghui; Zhang, Yi; Wang, Qiuping [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230026 (China); Wan, Baonian [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230026 (China)

2015-10-15

A toroidal charge exchange recombination spectroscopy (CXRS) diagnostic, on the basis of a heating neutral beam injector (NBI), is constructed on EAST tokamak. Simulation of Spectra (SOS) code is used to design and evaluate the diagnostic performance. 30 spatial channels work simultaneously in recent experiment, which covers a radial region from 1.55 m to 2.30 m in the cross section. The CXRS has a radial resolution of 1–3.5 cm from core to edge. The acquisition time is typically 10 ms, limited by the poor photon statistics. The diagnostic can observe not only the normal C{sup 5+} emission line at 529.1 nm but also any interested wavelength in the range of 400–700 nm. In this work, a brief overview on the R&D and the instrument performance for the toroidal CXRS diagnostic is described, together with first results.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

International Nuclear Information System (INIS)

Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

2015-01-01

Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

Propagation and absorption of the lower hybrid wave in a non-axisymmetric tokamak

International Nuclear Information System (INIS)

Arslanbekov, R.; Peysson, Y.; Basiuk, V.; Carrasco, J.; Hoang, G.T.; Litaudon, X.; Moreau, D.; Bizarro, J.P.; Ferreira, J.S.

1995-01-01

Studies have been carried out to assess how the LH wave propagation and absorption are affected by the magnetic ripple that is due to the finite number of coils used to create the toroidal field. It has been shown that the discreteness of the toroidal-field system may significantly alter the picture of LH wave propagation. It has been demonstrated that, for parameters of practical interest, magnetic ripple may induce stochastic behaviour in the ray dynamics. This work was extended to assess the effects of magnetic ripple on LH wave dynamics in a toroidal geometry, when both poloidal and toroidal inhomogeneities are present. The study is carried out for the Tore Supra tokamak. (K.A.) 7 refs.; 4 figs

Next generation toroidal devices

International Nuclear Information System (INIS)

Yoshikawa, Shoichi

1998-10-01

A general survey of the possible approach for the next generation toroidal devices was made. Either surprisingly or obviously (depending on one's view), the technical constraints along with the scientific considerations lead to a fairly limited set of systems for the most favorable approach for the next generation devices. Specifically if the magnetic field strength of 5 T or above is to be created by superconducting coils, it imposes minimum in the aspect ratio for the tokamak which is slightly higher than contemplated now for ITER design. The similar technical constraints make the minimum linear size of a stellarator large. Scientifically, it is indicated that a tokamak of 1.5 times in the linear dimension should be able to produce economically, especially if a hybrid reactor is allowed. For the next stellarator, it is strongly suggested that some kind of helical axis is necessary both for the (almost) absolute confinement of high energy particles and high stability and equilibrium beta limits. The author still favors a heliac most. Although it may not have been clearly stated in the main text, the stability afforded by the shearless layer may be exploited fully in a stellarator. (author)

Next generation toroidal devices

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Shoichi [Princeton Plasma Physics Lab., Princeton Univ., NJ (United States)

1998-10-01

A general survey of the possible approach for the next generation toroidal devices was made. Either surprisingly or obviously (depending on one`s view), the technical constraints along with the scientific considerations lead to a fairly limited set of systems for the most favorable approach for the next generation devices. Specifically if the magnetic field strength of 5 T or above is to be created by superconducting coils, it imposes minimum in the aspect ratio for the tokamak which is slightly higher than contemplated now for ITER design. The similar technical constraints make the minimum linear size of a stellarator large. Scientifically, it is indicated that a tokamak of 1.5 times in the linear dimension should be able to produce economically, especially if a hybrid reactor is allowed. For the next stellarator, it is strongly suggested that some kind of helical axis is necessary both for the (almost) absolute confinement of high energy particles and high stability and equilibrium beta limits. The author still favors a heliac most. Although it may not have been clearly stated in the main text, the stability afforded by the shearless layer may be exploited fully in a stellarator. (author)

Dynamics and feedback control of plasma equilibrium position in a tokamak

International Nuclear Information System (INIS)

Burenko, O.

1983-01-01

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

Periodic disruptions in the MT-1 tokamak

International Nuclear Information System (INIS)

Zoletnik, S.

1988-11-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

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

International Nuclear Information System (INIS)

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

1996-10-01

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

Turbulent transport of toroidal angular momentum in low flow gyrokinetics

International Nuclear Information System (INIS)

Parra, Felix I; Catto, Peter J

2010-01-01

We derive a self-consistent equation for the turbulent transport of toroidal angular momentum in tokamaks in the low flow ordering that only requires solving gyrokinetic Fokker-Planck and quasineutrality equations correct to second order in an expansion on the gyroradius over scale length. We also show that according to our orderings the long wavelength toroidal rotation and the long wavelength radial electric field satisfy the neoclassical relation that gives the toroidal rotation as a function of the radial electric field and the radial gradients of pressure and temperature. Thus, the radial electric field can be solved for once the toroidal rotation is calculated from the transport of toroidal angular momentum. Unfortunately, even though this methodology only requires a gyrokinetic model correct to second order in gyroradius over scale length, current gyrokinetic simulations are only valid to first order. To overcome this difficulty, we exploit the smallish ratio B p /B, where B is the total magnetic field and B p is its poloidal component. When B p /B is small, the usual first order gyrokinetic equation provides solutions that are accurate enough to employ for our expression for the transport of toroidal angular momentum. We show that current δf and full f simulations only need small corrections to achieve this accuracy. Full f simulations, however, are still unable to determine the long wavelength, radial electric field from the quasineutrality equation.

Free-boundary toroidal Alfvén eigenmodes

Science.gov (United States)

Chen, Eugene Y.; Berk, H. L.; Breizman, B.; Zheng, L. J.

2011-05-01

A numerical study is presented for the n = 1 free-boundary toroidal Alfvén eigenmodes (TAE) in tokamaks, which shows that there is considerable sensitivity of n = 1 modes to the position of the conducting wall. An additional branch of the TAE is shown to emerge from the upper continuum as the ratio of conducting wall radius to plasma radius increases. Such phenomena arise in plasma equilibria with both circular and shaped cross sections, where the shaped profile studied here is similar to that found in Alcator C-Mod.

Curvature-induced electrostatic drift modes in a toroidal plasma

International Nuclear Information System (INIS)

Venema, M.

1985-01-01

This thesis deals with a number of problems in the theory of linear stability of a hot, fully ionized plasma immersed in a strong magnetic field. The most widely used system to magnetically confine a plasma is the tokamak. This is a toroidal, current carrying device with a strong, externally imposed, magnetic field. The author discusses the linear theory of unstable, low-frequency waves in the gradient region, restricted to electrostatic waves. In that case the resulting radial fluxes of particles and energy are due to electric cross-field drifts. In the presence of magnetic fluctuations and small-scale reconnection phenomena, radial transport could also be predominantly along field lines. At present, it is not clear which of the two mechanisms is the dominant feature of the observed anomalous transport. First, the author introduces the theory of drift waves in toroidal geometry. Next, the electrostratic drift modes in toroidal geometry (weakly collisional regime), the equations for low-frequency waves in the strongly collisional regime and the electrostatic drift modes (strongly collisional regime) are discussed. (Auth.)

A tokamak with nearly uniform coil stress based on virial theorem

International Nuclear Information System (INIS)

Tsutsui, H.

2002-01-01

A novel tokamak concept with a new type of toroidal field (TF) coils and a central solenoid (CS) whose stress is much reduced to a theoretical limit determined by the virial theorem has been devised. Recently, we had developed a tokamak with force-balanced coils (FBCs) which are multi-pole helical hybrid coils combining TF coils and a CS coil. The combination reduces the net electromagnetic force in the direction of major radius. In this work, we have extended the FBC concept using the virial theorem. High-field coils should accordingly have same averaged principal stresses in all directions, whereas conventional FBC reduces stress in the toroidal direction only. Using a shell model, we have obtained the poloidal rotation number of helical coils which satisfy the uniform stress condition, and named the coil as virial-limited coil (VLC). VLC with circular cross section of aspect ratio A=2 reduces maximum stress to 60% compared with that of TF coils. In order to prove the advantage of VLC concept, we have designed a small VLC tokamak Todoroki-II. The plasma discharge in Todoroki-II will be presented. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1979-12-01

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

Shielding of External Magnetic Perturbations By Torque In Rotating Tokamak Plasmas

International Nuclear Information System (INIS)

Park, Jong-Kyu; Boozer, Allen H.; Menard, Jonathan E.; Gerhardt, Stefan P.; Sabbagh, Steve A.

2009-01-01

The imposition of a nonaxisymmetric magnetic perturbation on a rotating tokamak plasma requires energy and toroidal torque. Fundamental electrodynamics implies that the torque is essentially limited and must be consistent with the external response of a plasma equilibrium (rvec f) = (rvec j) x (rvec B). Here magnetic measurements on National Spherical Torus eXperiment (NSTX) device are used to derive the energy and the torque, and these empirical evaluations are compared with theoretical calculations based on perturbed scalar pressure equilibria (rvec f) = (rvec (del))p coupled with the theory of nonambipolar transport. The measurement and the theory are consistent within acceptable uncertainties, but can be largely inconsistent when the torque is comparable to the energy. This is expected since the currents associated with the torque are ignored in scalar pressure equilibria, but these currents tend to shield the perturbation.

A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.

1987-01-01

Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

Effect of continuous eigenvalue spectrum on plasma transport in toroidal systems

International Nuclear Information System (INIS)

Yamagishi, Tomejiro

1993-03-01

The effect of the continuous eigenvalue of the Vlasov equation on the cross field ion thermal flux is investigated. The continuum contribution due to the toroidal drift resonance is found to play an important role in ion transport particularly near the edge, which may apply to the interpretation of the sharp increase of ion heat conductivity near the periphery observed in large tokamaks. (author)

Toroidal Simulations of Sawteeth with Diamagnetic Effects

Science.gov (United States)

Beidler, Matthew; Cassak, Paul; Jardin, Stephen

2014-10-01

The sawtooth crash in tokamaks limits the core temperature, adversely impacts confinement, and seeds disruptions. Adequate knowledge of the physics governing the sawtooth crash and a predictive capability of its ramifications has been elusive, including an understanding of incomplete reconnection, i.e., why sawteeth often cease prematurely before processing all available magnetic flux. There is an indication that diamagnetic suppression could play an important role in this phenomenon. While computational tools to study toroidal plasmas have existed for some time, extended-MHD physics have only recently been integrated. Interestingly, incomplete reconnection has been observed in simulations when diamagnetic effects are present. In the current study, we employ the three-dimensional, extended-MHD code M3D-C1 to study the sawtooth crash in a toroidal geometry. In particular, we describe how magnetic reconnection at the q = 1 rational surface evolves when self-consistently increasing diamagnetic effects are present. We also explore how the termination of reconnection may lead to core-relaxing ideal-MHD instabilities.

The implementation of a toroidal limiter model into the gyrokinetic code ELMFIRE

Energy Technology Data Exchange (ETDEWEB)

Leerink, S.; Janhunen, S.J.; Kiviniemi, T.P.; Nora, M. [Euratom-Tekes Association, Helsinki University of Technology (Finland); Heikkinen, J.A. [Euratom-Tekes Association, VTT, P.O. Box 1000, FI-02044 VTT (Finland); Ogando, F. [Universidad Nacional de Educacion a Distancia, Madrid (Spain)

2008-03-15

The ELMFIRE full nonlinear gyrokinetic simulation code has been developed for calculations of plasma evolution and dynamics of turbulence in tokamak geometry. The code is applicable for calculations of strong perturbations in particle distribution function, rapid transients and steep gradients in plasma. Benchmarking against experimental reflectometry data from the FT2 tokamak is being discussed and in this paper a model for comparison and studying poloidal velocity is presented. To make the ELMFIRE code suitable for scrape-off layer simulations a simplified toroidal limiter model has been implemented. The model is be discussed and first results are presented. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Stationary shear flows in CGL anisotropic toroidal plasmas

International Nuclear Information System (INIS)

Pastukhov, V.P.; Ilgisonis, V.I.

1996-01-01

Recently a general structure of stationary shear flows in toroidal plasmas was obtained in the frame of ideal isotropic-pressure MHD model. The structure of the stationary plasma flows was shown to be determined by a hidden symmetry of MHD equations inherent in the toroidal systems with nested magnetic surfaces. However, the characteristic frequencies of the stationary plasma motion can considerably exceed the collisional frequencies in real plasma experiments. In this case the CGL collisionless MHD model seems to be more adequate than the simplified isotropic-pressure MHD model to describe the stationary plasma flows. In this paper we have generalized our approach to analyze the stationary plasma flows in the frame of the collisionless CGL model. We have found again that the hidden symmetry inherent in the toroidal topology results in two integral invariants which depend on two independent surface functions. The structure of stationary flows for CGL model is still the same as for isotropic MHD, however, the pressure tensor components satisfy a appreciably modifies the steady state force-balance equation. These results are applied to analyze the generalized equilibrium in axisymmetric (tokamak-like) magnetic confinement systems

Elastic stability and vibration of toroidal magnets for fusion reactors. Final report

International Nuclear Information System (INIS)

Moon, F.C.; Swanson, C.

1975-09-01

The vibration and elastic stability of a set of discrete superconducting toroidal field magnets arranged to form a ''bumpy'' torus is examined. The mutual destabilizing magnetic forces between magnet pairs are calculated using a numerical differential inductance technique. It is shown that the mutual attractive magnetic forces can produce elastic buckling of the entire toroidal set. The vibration modes of the set are also found as functions of the coil current. The response of the set of magnets to an earthquake type motion of the toroidal base is calculated. The calculations have been incorporated in a computer code which accompanies the report. Measurements are made of the lateral stiffness of a flexible, planar, superconducting coil between two rigid coils in series. These tests show a dramatic decrease in the natural bending frequency with subsequent elastic instability or ''buckling'' at a critical value of the current in the coils. These observations support a magnetoelastic analysis which shows that proposed designs, of toroidal field coils for Tokamak fusion reactors, have insufficient lateral support for mechanical stability of the magnets

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

International Nuclear Information System (INIS)

In, Sang Ryul; Yoon, Byung Joo.

1996-11-01

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

Modelling of the toroidal asymmetry of poloidal halo currents in conducting structures

International Nuclear Information System (INIS)

Pomphrey, N.; Bialek, J.M.; Part, W.

1998-01-01

During plasma disruptions, substantial toroidal and poloidal eddy currents are generated in the vacuum vessel and other plasma facing conducting structures. Eddy currents that conduct charge through paths which close through the plasma periphery are called halo currents, and these can be of substantial magnitude. Of particular concern for tokamak design and operation is the observed toroidal asymmetry of the halo current distribution: such an asymmetric distribution leads to problematic non-uniform forces on the conducting structures. The premise is adopted that the source of toroidal asymmetry is the plasma deformation resulting from the non-linear external kink instability that develops during the current quench phase of a disruption. A simple model is presented of the kinked plasma that allows an analytic calculation of the dependence of the toroidal peaking factor (TPF) on the ratio of the halo current to the total toroidal plasma current, I h /I p . Expressions for the TPF as a function of I h /I p are derived for m/n=2/1 and m/n=1/1 helical instabilities. The expressions depend on a single parameter, which measures the amplitude of the saturated state of the kink instability. A comparison with disruption data from experiments shows good agreement. Numerical experiments that simulate non-linear external kinks provide guidance on the values expected for the saturated amplitude. It is proposed that a simple plasma halo model is adequate for assessing the engineering impact of asymmetric halo currents, since the force distribution on the conducting structures depends mainly on the 'resistive distribution' of the eddy currents. A brief description is given of an electromagnetics code that calculates the time development of eddy currents in conducting structures, and the code is applied to two halo current disruption scenarios. These are used to emphasize the importance of having an accurate eddy current calculation to correctly estimate the engineering impact of

Relativistic runaway electrons in tokamak plasmas

International Nuclear Information System (INIS)

Jaspers, R.E.

1995-01-01

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

Spherical tokamak power plant design issues

International Nuclear Information System (INIS)

Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

2000-01-01

The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

Shear Alfven waves in tokamaks

International Nuclear Information System (INIS)

Kieras, C.E.

1982-12-01

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

Equations for the non linear evolution of the resistive tearing modes in toroidal plasmas

International Nuclear Information System (INIS)

Edery, D.; Pellat, R.; Soule, J.L.

1979-09-01

Following the tokamak ordering, we simplify the resistive MHD equations in toroidal geometry. We obtain a closed system of non linear equations for two scalar potentials of the magnetic and velocity fields and for plasma density and temperature. If we expand these equations in the inverse of aspect ratio they are exact to the two first orders. Our formalism should correctly describe the mode coupling by curvature effects /1/ and the toroidal displacement of magnetic surfaces /2/. It provides a natural extension of the well known cylindrical model /3/ and is now being solved on computer

Development of 3D ferromagnetic model of tokamak core withstrong toroidal asymmetry

Czech Academy of Sciences Publication Activity Database

Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Pánek, Radomír

96-97, October (2015), s. 302-305 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * ferromagnetic core * model of ferromagnet * integral method * tokamak GOLEM Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615002100

Study of runaway electrons using the conditional average sampling method in the Damavand tokamak

Energy Technology Data Exchange (ETDEWEB)

Pourshahab, B., E-mail: bpourshahab@gmail.com [University of Isfahan, Department of Nuclear Engineering, Faculty of Advance Sciences and Technologies (Iran, Islamic Republic of); Sadighzadeh, A. [Nuclear Science and Technology Research Institute, Plasma Physics and Nuclear Fusion Research School (Iran, Islamic Republic of); Abdi, M. R., E-mail: r.abdi@phys.ui.ac.ir [University of Isfahan, Department of Physics, Faculty of Science (Iran, Islamic Republic of); Rasouli, C. [Nuclear Science and Technology Research Institute, Plasma Physics and Nuclear Fusion Research School (Iran, Islamic Republic of)

2017-03-15

Some experiments for studying the runaway electron (RE) effects have been performed using the poloidal magnetic probes system installed around the plasma column in the Damavand tokamak. In these experiments, the so-called runaway-dominated discharges were considered in which the main part of the plasma current is carried by REs. The induced magnetic effects on the poloidal pickup coils signals are observed simultaneously with the Parail–Pogutse instability moments for REs and hard X-ray bursts. The output signals of all diagnostic systems enter the data acquisition system with 2 Msample/(s channel) sampling rate. The temporal evolution of the diagnostic signals is analyzed by the conditional average sampling (CAS) technique. The CASed profiles indicate RE collisions with the high-field-side plasma facing components at the instability moments. The investigation has been carried out for two discharge modes—low-toroidal-field (LTF) and high-toroidal-field (HTF) ones—related to both up and down limits of the toroidal magnetic field in the Damavand tokamak and their comparison has shown that the RE confinement is better in HTF discharges.

Simulations of toroidal Alfvén eigenmode excited by fast ions on the Experimental Advanced Superconducting Tokamak

Science.gov (United States)

Pei, Youbin; Xiang, Nong; Shen, Wei; Hu, Youjun; Todo, Y.; Zhou, Deng; Huang, Juan

2018-05-01

Kinetic-MagnetoHydroDynamic (MHD) hybrid simulations are carried out to study fast ion driven toroidal Alfvén eigenmodes (TAEs) on the Experimental Advanced Superconducting Tokamak (EAST). The first part of this article presents the linear benchmark between two kinetic-MHD codes, namely MEGA and M3D-K, based on a realistic EAST equilibrium. Parameter scans show that the frequency and the growth rate of the TAE given by the two codes agree with each other. The second part of this article discusses the resonance interaction between the TAE and fast ions simulated by the MEGA code. The results show that the TAE exchanges energy with the co-current passing particles with the parallel velocity |v∥ | ≈VA 0/3 or |v∥ | ≈VA 0/5 , where VA 0 is the Alfvén speed on the magnetic axis. The TAE destabilized by the counter-current passing ions is also analyzed and found to have a much smaller growth rate than the co-current ions driven TAE. One of the reasons for this is found to be that the overlapping region of the TAE spatial location and the counter-current ion orbits is narrow, and thus the wave-particle energy exchange is not efficient.

Technology issues for decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Spampinato, P.T.; Walton, G.R.

1994-01-01

The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community

Stabilization of external kink modes in a tokamak with rotating plasma

International Nuclear Information System (INIS)

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

1995-01-01

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

Maximum attainable power density and wall load in tokamaks underlying reactor relevant constraints

International Nuclear Information System (INIS)

Borrass, K.; Buende, R.

1979-09-01

The characteristic data of tokamaks optimized with respect to their power density or wall load are determined. Reactor relevant constraints are imposed, such as a fixed plant net power output, a fixed blanket thickness and the dependence of the maximum toroidal field on the geometry and conductor material. The impact of finite burn times is considered. Various scaling laws of the toroidal beta with the aspect ratio are discussed. (orig.) 891 GG/orig. 892 RDG [de

Runaway electrons in toroidal discharges

International Nuclear Information System (INIS)

Knoepfel, H.

1979-01-01

Experimental and theoretical studies of runaway electrons in toroidal devices are reviewed here, with particular reference to tokamaks. The complex phenomenology of runaway effects, which have been the subject of research for the past twenty years, is organized within the framework of a number of physical models. The mechanisms and rates for runaway production are discussed first, followed by sections on runaway-driven kinetic relaxation processes and runaway orbit confinement. Next, the equilibrium and stability of runaway-dominated discharges are reviewed. Models for runaway production at early times in the discharge and the scaling of runaway phenomena to larger devices are also discussed. Finally, detection techniques and possible applications of runaways are mentioned. (author)

Turbulence and abnormal transport in tokamak plasmas

International Nuclear Information System (INIS)

Garbet, X.

1988-09-01

Microinstabilities in linear and nonlinear tokamak plasmas were studied. A variational method based on the existence of a system of angular variables and action for the charged particles in the magnetic configuration of a tokamak is described. The corresponding functional, extremal in relation to the fluctuating electromagnetic field, is calculated analytically, taking into account the effects of the toroidal geometry. A numerical code, TORRID, was derived from these principles and the main instabilities, especially ion instabilities and microtearing, were studied linearly. Nonlinear methods were also applied to microtearing. Quasi-linear transport coefficients are derived from a principle of minimum entropy production. Thermal ionic conductivity and viscosity are calculated for an ionic turbulence [fr

Linear and nonlinear kinetic-stability studies in tokamaks

International Nuclear Information System (INIS)

Tang, W.M.; Chance, M.S.; Chen, L.; Krommes, J.A.; Lee, W.W.; Rewoldt, G.

1982-09-01

This paper presents results of theoretical investigations on important linear kinetic properties of low frequency instabilities in toroidal systems and on nonlinear processes which could significantly influence their impact on anomalous transport. Analytical and numerical methods and also particle simulations have been employed to carry out these studies. In particular, the following subjects are considered: (1) linear stability analysis of kinetic instabilities for realistic tokamak equilibria and the application of such calculations to the PDX and PLT tokamak experiments including the influence of a hot beam-ion component; (2) determination of nonlinearly saturated, statistically steady states of three interacting drift modes; and (3) gyrokinetic particle simulation of drift instabilities

Resonant MHD modes with toroidal coupling

International Nuclear Information System (INIS)

Connor, J.W.; Hastie, R.J.; Taylor, J.B.

1990-07-01

This is part 2 of a study of resonant perturbations, such as resistive tearing and ballooning modes, in a torus. These are described by marginal ideal mhd equations in the regions between resonant surfaces; matching across these surfaces provides the dispersion relation. In part 1 we described how all the necessary information from the ideal mhd calculations could be represented by a so-called E-matrix. We also described the calculation of this E-matrix for tearing modes (even parity in perturbed magnetic field) in a large aspect ratio torus. There the toroidal modes comprise coupled cylinder tearing modes and the E-matrix is a generalization of the familiar Δ' quantity in a cylinder. In the present paper we discuss resistive ballooning, or twisting-modes, which have odd-parity in perturbed magnetic field. We show that, unlike the tearing modes, these odd-parity modes are instrinsically toroidal and are not directly related to the odd-parity modes in a cylinder. This is evident from the analysis of the high-n limit in ballooning-space, where a transition from a stable Δ' to an unstable Δ' occurs for the twisting mode when the ballooning effect exceeds the interchange effect, which can occur even at large aspect ratio (as in a tokamak). Analysis of the high-n limit in coordinate space, rather than ballooning space, clarifies this singular behaviour and indicates how one may define twisting-mode Δ'. It also yields a prescription for treating low-n twisting modes and a method for calculating an E-matrix for resistive ballooning modes in a large aspect ratio tokamak. The elements of this matrix are given in terms of cylindrical tearing mode solutions

Equilibrium, confinement and stability of runaway electrons in tokamaks

International Nuclear Information System (INIS)

Spong, D.A.

1976-03-01

Some of the ramifications of the runaway population in tokamak experiments are investigated. Consideration is given both to the normal operating regime of tokamaks where only a small fraction of high energy runaways are present and to the strong runaway regime where runaways are thought to carry a significant portion of the toroidal current. In particular, the areas to be examined are the modeling of strong runaway discharges, single particle orbit characteristics of runaways, macroscopic beam-plasma equilibria, and stability against kink modes. A simple one-dimensional, time-dependent model has been constructed in relation to strong runaway discharges. Single particle orbits are analyzed in relation to both the strong runaway regime and the weak regime. The effects of vector E x vector B drifts are first considered in strong runaway discharges and are found to lead to a slow inward shrinkage of the beam. Macroscopic beam-plasma equilibria are treated assuming a pressureless relativistic beam with inertia and using an ideal MHD approximation for the plasma. The stability of a toroidal relativistic beam against kink perturbations is examined using several models

Numerical investigation of non-perturbative kinetic effects of energetic particles on toroidicity-induced Alfvén eigenmodes in tokamaks and stellarators

International Nuclear Information System (INIS)

Slaby, Christoph; Könies, Axel; Kleiber, Ralf

2016-01-01

The resonant interaction of shear Alfvén waves with energetic particles is investigated numerically in tokamak and stellarator geometry using a non-perturbative MHD-kinetic hybrid approach. The focus lies on toroidicity-induced Alfvén eigenmodes (TAEs), which are most easily destabilized by a fast-particle population in fusion plasmas. While the background plasma is treated within the framework of an ideal-MHD theory, the drive of the fast particles, as well as Landau damping of the background plasma, is modelled using the drift-kinetic Vlasov equation without collisions. Building on analytical theory, a fast numerical tool, STAE-K, has been developed to solve the resulting eigenvalue problem using a Riccati shooting method. The code, which can be used for parameter scans, is applied to tokamaks and the stellarator Wendelstein 7-X. High energetic-ion pressure leads to large growth rates of the TAEs and to their conversion into kinetically modified TAEs and kinetic Alfvén waves via continuum interaction. To better understand the physics of this conversion mechanism, the connections between TAEs and the shear Alfvén wave continuum are examined. It is shown that, when energetic particles are present, the continuum deforms substantially and the TAE frequency can leave the continuum gap. The interaction of the TAE with the continuum leads to singularities in the eigenfunctions. To further advance the physical model and also to eliminate the MHD continuum together with the singularities in the eigenfunctions, a fourth-order term connected to radiative damping has been included. The radiative damping term is connected to non-ideal effects of the bulk plasma and introduces higher-order derivatives to the model. Thus, it has the potential to substantially change the nature of the solution. For the first time, the fast-particle drive, Landau damping, continuum damping, and radiative damping have been modelled together in tokamak- as well as in stellarator geometry.

Numerical investigation of non-perturbative kinetic effects of energetic particles on toroidicity-induced Alfvén eigenmodes in tokamaks and stellarators

Energy Technology Data Exchange (ETDEWEB)

Slaby, Christoph; Könies, Axel; Kleiber, Ralf [Max-Planck-Institut für Plasmaphysik, D-17491 Greifswald (Germany)

2016-09-15

The resonant interaction of shear Alfvén waves with energetic particles is investigated numerically in tokamak and stellarator geometry using a non-perturbative MHD-kinetic hybrid approach. The focus lies on toroidicity-induced Alfvén eigenmodes (TAEs), which are most easily destabilized by a fast-particle population in fusion plasmas. While the background plasma is treated within the framework of an ideal-MHD theory, the drive of the fast particles, as well as Landau damping of the background plasma, is modelled using the drift-kinetic Vlasov equation without collisions. Building on analytical theory, a fast numerical tool, STAE-K, has been developed to solve the resulting eigenvalue problem using a Riccati shooting method. The code, which can be used for parameter scans, is applied to tokamaks and the stellarator Wendelstein 7-X. High energetic-ion pressure leads to large growth rates of the TAEs and to their conversion into kinetically modified TAEs and kinetic Alfvén waves via continuum interaction. To better understand the physics of this conversion mechanism, the connections between TAEs and the shear Alfvén wave continuum are examined. It is shown that, when energetic particles are present, the continuum deforms substantially and the TAE frequency can leave the continuum gap. The interaction of the TAE with the continuum leads to singularities in the eigenfunctions. To further advance the physical model and also to eliminate the MHD continuum together with the singularities in the eigenfunctions, a fourth-order term connected to radiative damping has been included. The radiative damping term is connected to non-ideal effects of the bulk plasma and introduces higher-order derivatives to the model. Thus, it has the potential to substantially change the nature of the solution. For the first time, the fast-particle drive, Landau damping, continuum damping, and radiative damping have been modelled together in tokamak- as well as in stellarator geometry.

Tokamak startup using point-source dc helicity injection.

Science.gov (United States)

Battaglia, D J; Bongard, M W; Fonck, R J; Redd, A J; Sontag, A C

2009-06-05

Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.

Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

International Nuclear Information System (INIS)

Lee, V.D.; Gohar, Y.

1985-01-01

A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of builtup, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

International Nuclear Information System (INIS)

Lee, V.D.; Gohar, Y.

1985-01-01

A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

Energetic particle destabilization of shear Alfven waves in stellarators and tokamaks

International Nuclear Information System (INIS)

Spong, D.A.; Carreras, B.A.; Hedrick, C.L.; Leboeuf, J.N.; Weller, A.

1994-01-01

An important issue for ignited devices is the resonant destabilization of shear Alfven waves by energetic populations. These instabilities have been observed in a variety of toroidal plasma experiments in recent years, including: beam-destabilized toroidal Alfven instabilities (TAE) in low magnetic field tokamaks, ICRF destabilized TAE's in higher field tokamaks, and global Alfven instabilities (GAE) in low shear stellarators. In addition, excitation and study of these modes is a significant goal of the TFIR-DT program and a component of the ITER physics tasks. The authors have developed a gyrofluid model which includes the wave-particle resonances necessary to excite such instabilities. The TAE linear mode structure is calculated nonperturbatively, including many of the relevant damping mechanisms, such as: continuum damping, non-ideal effects (ion FLR and electron collisionality), and ion/electron Landau damping. This model has been applied to both linear and nonlinear regimes for a range of experimental cases using measured profiles

Minimum dimension of an ITER like Tokamak with a given Q

Energy Technology Data Exchange (ETDEWEB)

Johner, J

2004-07-01

The minimum dimension of an ITER like tokamak with a given amplification factor Q is calculated for two values of the maximum magnetic field in the superconducting toroidal field coils. For ITERH-98P(y,2) scaling of the energy confinement time, it is shown that for a sufficiently large tokamak, the maximum Q is obtained for the operating point situated both at the maximum density and at the minimum margin with respect to the H-L transition. We have shown that increasing the maximum magnetic field in the toroidal field coils from the present 11.8 T to 16 T would result in a strong reduction of the machine size but has practically no effect on the fusion power. Values obtained for {beta}{sub N} are found to be below 2. Peak fluxes on the divertor plates with an ITER like divertor and a multi-machine expression for the power radiated in the plasma mantle, are below 10 MW/m{sup 2}.

Nonlinear hybrid simulation of toroidicity-induced alfven eigenmode

International Nuclear Information System (INIS)

Fu, G.Y.; Park, W.

1994-11-01

Gyrokinetic/Magnetohydrodynamics hybrid simulations have been carried out using MH3D-K code to study the nonlinear saturation of the toroidicity-induced Alfven eigenmode driven by energetic particles in a tokamak plasma. It is shown that the wave particle trapping is the nonlinear saturation mechanism for the parameters considered. The corresponding density profile flattening of hot particles is observed. The saturation amplitude is proportional to the square of linear growth rate. In addition to TAE modes, a new n = 1, m = 0 global Alfven eigenmode is shown to be excited by the energetic particles

General Atomic's superconducting toroidal field coil concept

International Nuclear Information System (INIS)

Alcorn, J.; Purcell, J.

1978-01-01

General Atomic's concept for a superconducting toroidal field coil is presented. The concept is generic for large tokamak devices, while a specific design is indicated for a 3.8 meter (major radius) ignition/burn machine. The concept utilizes bath cooled NbTi conductor to generate a peak field of 10 tesla at 4.2 K. The design is simple and straightforward, requires a minimum of developmental effort, and draws extensively upon the perspective of past experience in the design and construction of large superconducting magnets for high energy physics. Thus, the primary emphasis is upon economy, reliability, and expeditious construction scheduling. (author)

The Tokamak Fusion Core Experiment studies

International Nuclear Information System (INIS)

Schmidt, J.A.; Sheffield, G.V.; Bushnell, C.

1985-01-01

The basic objective of the next major step in the US fusion programme has been defined as the achievement of ignition and long pulse equilibrium burn of a fusion plasma in the Tokamak Fusion Core Experiment (TFCX) device. Preconceptual design studies have seen completion of four candidate versions to provide the comparative information needed to narrow down the range of TFCX options before proceeding to the conceptual design phase. All four designs share the same objective and conform to common physics, engineering and costing criteria. The four base options considered differed mainly in the toroidal field coil design, two employing superconducting coils and the other two copper coils. In each case (copper and superconducting), one relatively conventional version was carried as well as a version employing more exotic toroidal field coil design assumptions. Sizes range from R=2.6 m for the smaller of the two copper versions to R=4.08 m for the larger superconducting option. In all cases, the plasma current was about 10 MA and the toroidal field about 4 T. (author)

Three-dimensional tokamak equilibria and stellarators with two-dimensional magnetic symmetry

International Nuclear Information System (INIS)

Garabedian, P.R.

1997-01-01

Three-dimensional computer codes have been developed to simulate equilibrium, stability and transport in tokamaks and stellarators. Bifurcated solutions of the tokamak problem suggest that three-dimensional effects may be more important than has generally been thought. Extensive calculations have led to the discovery of a stellarator configuration with just two field periods and with aspect ratio 3.2 that has a magnetic field spectrum B mn with toroidal symmetry. Numerical studies of equilibrium, stability and transport for this new device, called the Modular Helias-like Heliac 2 (MHH2), will be presented. (author)

On the choice of toroidal magnetic field for thermonuclear tokamaks

International Nuclear Information System (INIS)

Segre, S.E.

1981-01-01

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

Sheared Rotation Effects on Kinetic Stability in Enhanced Confinement Tokamak Plasmas, and Nonlinear Dynamics of Fluctuations and Flows in Axisymmetric Plasmas

International Nuclear Information System (INIS)

Beer, M.A.; Chance, M.S.; Hahm, T.S.; Lin, Z.; Rewoldt, G.; Tang, W.M.

1997-01-01

Sheared rotation dynamics are widely believed to have signficant influence on experimentally observed confinement transitions in advanced operating modes in major tokamak experiments, such as the Tokamak Fusion Test Reactor (TFTR) [D.J. Grove and D.M. Meade, Nuclear Fusion 25, 1167 (1985)], with reversed magnetic shear regions in the plasma interior. The high-n toroidal drift modes destabilized by the combined effects of ion temperature gradients and trapped particles in toroidal geometry can be strongly affected by radially sheared toroidal and poloidal plasma rotation. In previous work with the FULL linear microinstability code, a simplified rotation model including only toroidal rotation was employed, and results were obtained. Here, a more complete rotation model, that includes contributions from toroidal and poloidal rotation and the ion pressure gradient to the total radial electric field, is used for a proper self-consistent treatment of this key problem. Relevant advanced operating mode cases for TFTR are presented. In addition, the complementary problem of the dynamics of fluctuation-driven E x B flow is investigated by an integrated program of gyrokinetic simulation in annulus geometry and gyrofluid simulation in flux tube geometry

Edge plasma diagnostics on Tore Supra tokamak

International Nuclear Information System (INIS)

Fujita, Junji

1991-01-01

From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)

Development and Operational Experiences of the JT-60U Tokamak and Power Supplies

International Nuclear Information System (INIS)

Hosogane, N.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Neyatani, Y.; Horiike, H.; Sakurai, S.; Masaki, K.; Yamamoto, M.; Kodama, K.; Sasajima, T.; Terakado, T.; Ohmori, S.; Ohmori, Y.; Okano, J.

2002-01-01

The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-60 in order to obtain high plasma current, large plasma volume, and highly elongated divertor configurations. All components inside the toroidal magnetic field coils, such as vacuum vessel, poloidal magnetic field coils, divertor, etc., were modified. Various technologies and ideas were introduced to develop these components; for example, a multi-arc double skin wall structure for the vacuum vessel and a functional poloidal magnetic field coil system with taps for obtaining various plasma configurations. Furthermore, boron-carbide coated carbon fiber composite (CFC) tiles were used as divertor tiles to reduce erosion of carbon-base tiles. Later, a semiclosed divertor with pumps, for which cryo-panels originally used for NBI units were converted, was installed in the replacement of the open divertor. These development and operational results provide data for future tokamaks. Major failures experienced in the long operational period of JT-60U, such as water leakage from the toroidal magnetic field coil, fracture of carbon tiles, and breakdown of a filter capacitor, are described. As a maintenance issue for tokamaks using deuterium fueling gas, a method for reducing radiation exposure of in-vessel workers is described

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

Science.gov (United States)

Crank, Willie; Ali, Halima; Punjabi, Alkesh

2009-11-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Conceptual Design of Alborz Tokamak Poloidal Coils System

Science.gov (United States)

Mardani, M.; Amrollahi, R.

2013-04-01

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.

On the parametric cyclotron heating of a toroidal plasma

International Nuclear Information System (INIS)

Golovanivsky, K.C.; Punithavelu, A.M.

1976-01-01

The possibility of heating the ionic component of a dense plasma at the parametric cyclotron resonance, using a section of the conducting toroidal chamber of a large scale Tokamak as a resonance cavity, is considered. It is suggested to use the mode TE 011 to overcome the difficulties with the penetration of HF fields into such a dense plasma. The experimental investigation of parametric cyclotron heating of electrons in a overdense plasma (n/nsub(cut off)=10 2 ) on such a model has given hopeful results

Magnetohydrodynamic stability of tokamak edge plasmas

International Nuclear Information System (INIS)

Connor, J.W.; Hastie, R.J.; Wilson, H.R.; Miller, R.L.

1998-01-01

A new formalism for analyzing the magnetohydrodynamic stability of a limiter tokamak edge plasma is developed. Two radially localized, high toroidal mode number n instabilities are studied in detail: a peeling mode and an edge ballooning mode. The peeling mode, driven by edge current density and stabilized by edge pressure gradient, has features which are consistent with several properties of tokamak behavior in the high confinement open-quotes Hclose quotes-mode of operation, and edge localized modes (or ELMs) in particular. The edge ballooning mode, driven by the pressure gradient, is identified; this penetrates ∼n 1/3 rational surfaces into the plasma (rather than ∼n 1/2 , expected from conventional ballooning mode theory). Furthermore, there exists a coupling between these two modes and this coupling provides a picture of the ELM cycle

Tore Supra. Basic design Tokamak system

International Nuclear Information System (INIS)

Aymar, R.; Bareyt, B.; Bon Mardion, G.

1980-10-01

This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

Kinetic global analysis of Alfven eigenmodes in toroidal plasmas

International Nuclear Information System (INIS)

Fukuyama, A.

2002-01-01

Systematic study on low to medium n (toroidal mode number) Alfven eigenmodes (AE) in tokamaks and helical systems is presented. Linear stability of AE in the presence of energetic ions was studied using the kinetic full-wave code TASK/WM.We have reproduced the destabilizing effect of toroidal co-rotation on TAE for JT-60U parameters. We have found the existence of reversed-shear-induced Alfven eigenmode (RSAE) which localizes near the q minimum in a reversed magnetic shear configuration. Two kinds of mode structures are identified for energetic particle mode (EPM) below the TAE frequency gap. The coupling to lower-frequency modes such as drift waves and MHD modes as well as the effect of trapped particles are also taken into account. For a helical plasma, the existence of GAE in the central region and TAE in the off-axis region was confirmed. (author)

MHD equilibrium of toroidal fusion plasma with stationary flows

International Nuclear Information System (INIS)

Galkowski, A.

1994-01-01

Non-linear ideal MHD equilibria in axisymmetric system with flows are examined, both in 1st and 2nd ellipticity regions. Evidence of the bifurcation of solutions is provided and numerical solutions of several problems in a tokamak geometry are given, exhibiting bifurcation phenomena. Relaxation of plasma in the presence of zero-order flows is studied in a realistic toroidal geometry. The field aligned flow allows equilibria with finite pressure gradient but with homogeneous temperature distribution. Numerical calculations have been performed for the 1st and 2nd ellipticity regimes of the extended Grad-Shafranov-Schlueter equation. Numerical technique, alternative to the well-known Grad's ADM methods has been proposed to deal with slow adiabatic evolution of toroidal plasma with flows. The equilibrium problem with prescribed adiabatic constraints may be solved by simultaneous calculations of flux surface geometry and original profile functions. (author). 178 refs, 37 figs, 5 tabs

Engineering feasibility of tight aspect ratio Tokamak (spherical torus) reactors

International Nuclear Information System (INIS)

Peng, Y-K.M.; Hicks, J.B.

1990-01-01

Engineering solutions are identified and analyzed for key high-power-density components of tight aspect ratio tokamak reactors (spherical torus reactors). The potentially extreme divertor heat loads can be reduced to about 3 MW/m 2 in expanded divertors using coils inside the demountable toroidal field coils. Given the long and narrow divertor channels, gaseous divertor targets become possible, which eliminate sputtering and increase the divertor life. The unshielded centre conductor post (CCP) of the toroidal field coil can be made of a single dispersion strengthened copper conductor cooled by high-velocity pressurized water to maintain acceptable copper temperature and strength. Damage and activation of the CCP at a neutron fluence of 10 MW-a/m 2 are also tolerable. Annual replacement of the centre post, the divertor assemblies and the blanket can be accomplished with vertical access for all torus components, which are modularized to reduce size and weight. The technical requirements of these solutions are shown to be comparable with, if not less demanding than, those estimated for conventional tokamak reactors. (author)

The effect of non-inductive current drive on tokamak transport

International Nuclear Information System (INIS)

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

2005-01-01

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

New results from the Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.K.; Ananiev, A.S.; Amoskov, V.M.

2003-01-01

New results from the Globus-M spherical tokamak are presented. High plasma current of 0.36 MA, high toroidal magnetic field of 0.55 T and other important plasma characteristics were achieved. Described are the operational space and plasma stability limits in the OH regime. The factors limiting operational space (MHD instabilities, runaway electrons, etc.) are discussed. New experiments on plasma fuelling are described. First results of experiments with a coaxial plasma gun injector are presented. Initial results of a plasma - wall interaction study are outlined. First results obtained with new diagnostic tools installed on the tokamak are presented. An auxiliary heating system test was performed. Preliminary results of simulations and experiments are given. (author)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Toroidal electron beam energy storage for controlled fusion

International Nuclear Information System (INIS)

Clark, W.; Korn, P.; Mondelli, A.; Rostoker, N.

1976-01-01

In the presence of an external magnetic field stable equilibria exist for an unneutralized electron beam with ν/γ >1. As a result, it is in principle, possible to store very large quantities of energy in relatively small volumes by confining an unneutralized electron beam in a Tokamak-like device. The energy is stored principally in the electrostatic and self-magnetic fields associated with the beam and is available for rapid heating of pellets for controlled fusion. The large electrostatic potential well in such a device would be sufficient to contain energetic alpha particles, thereby reducing reactor wall bombardment. This approach also avoids plasma loss and wall bombardment by charge exchange neutrals. The conceptual design of an electrostatic Tokamak fusion reactor (ETFR) is discussed. A small toroidal device (the STP machine) has been constructed to test the principles involved. Preliminary experiments on this device have produced electron densities approximately 10% of those required in a reactor

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

International Nuclear Information System (INIS)

Ehst, D.A.

1981-02-01

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

The design of the KSTAR tokamak

International Nuclear Information System (INIS)

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

1999-01-01

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

The basics of spherical tokamaks and progress in European research

International Nuclear Information System (INIS)

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

2003-01-01

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

A tokamak reactor with servicing capability

International Nuclear Information System (INIS)

Mitchell, J.T.D.; Hollis, A.

1976-01-01

A conceptual design for a Tokamak reactor with practical facilities for the regular replacement of blanket components after the inevitable damage from neutron irradiation, and fatigue is described. This essential facility has been largely ignored in published fusion reactor designs. One exception is the inertially-confined Saturn proposal. Tokamak and other toroidal closed-line systems have very complex geometries and sub-system requirements, which result in blanket servicing being a very difficult problem. In the concept described the magnet shield is divided into two structures - an outer permanent one with access doors and an inner shield, part of and supporting the blanket inside. Servicing access is horizontally between the toroidal magnet coils, after moving some outer poloidal magnet coils. The reactor, reactor hall, workshops and remote-handling facilities are described, and the servicing requirements discussed. The important servicing operation is the remote replacement of radiation damaged blanket and shield - divided in this design into 20 sectors, each weighing 75-100 tons and 11-12 metres high. Analysis of the operation indicates that if one sector can be replaced during a single weekend - i.e. a period of low power demand - then the annual reactor-generator availability allowing as well for the general plant servicing should be >0.9. This level of availability should meet the requirements of generating authorities but the facilities, equipment and workshops necessary may be complex and expensive

Steady State Advanced Tokamak (SSAT): The mission and the machine

International Nuclear Information System (INIS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

Tokamak equilibria with non-parallel flow in a triangularity-deformed axisymmetric toroidal coordinate system

Directory of Open Access Journals (Sweden)

Ap Kuiroukidis

2018-01-01

Full Text Available We consider a generalized Grad–Shafranov equation (GGSE in a triangularity-deformed axisymmetric toroidal coordinate system and solve it numerically for the generic case of ITER-like and JET-like equilibria with non-parallel flow. It turns out that increase of the triangularity improves confinement by leading to larger values of the toroidal beta and the safety factor. This result is supported by the application of a criterion for linear stability valid for equilibria with flow parallel to the magnetic field. Also, the parallel flow has a weaker stabilizing effect.

Impact of major design parameters on the economics of Tokamak power plants

International Nuclear Information System (INIS)

Abdou, M.A.; Ehst, D.; Maroni, V.; Stacey, W.M. Jr.

1977-11-01

A parametric systems studies program is now in an active stage at Argonne National Laboratory. This paper presents a summary of results from this systems analysis effort. The impact of major design parameters on the economics of tokamak power plants is examined. The major parameters considered are: (1) the plant power rating; (2) toroidal-field strength; (3) plasma β/sub t/; (4) aspect ratio; (5) plasma elongation; (6) inner blanket/shield thickness; and (7) neutron wall load. The performance characteristics and economics of tokamak power plants are also compared for two structural materials

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Experimental studies of thermal and non-thermal electron cyclotron phenomena in tokamaks

International Nuclear Information System (INIS)

McDermott, F.S.

1984-12-01

A direct measurement of wave absorption in the ISX-B tokamak at the second harmonic of the electron cyclotron frequency is reported. Measurements of the absorption of a wave polarized in the extraordinary mode and propagating perpendicular to the toroidal magnetic field are in agreement with the absorption predicted by the linearized Vlasov equation for a thermal plasma. Agreement is found both for an analytic approximation to the wave absorption and for a numerical simulation of ray propagation in toroidal geometry. Observations are also reported on a non-linear, three-wave interaction process occurring during high power electron cyclotron resonance heating in the Versator II tokamak. The measured spectra and the threshold power are consistent with a model in which the incident power in the extraordinary mode of polarization decays at the upper hybrid resonance layer into a lower hybrid wave and an electron Bernstein wave. Finally, measurements of non-thermal emission at the second harmonic of the electron cyclotron frequency and below the electron plasma frequency are reported from low density, non-Maxwellian plasma in the Versator II tokamak. The emission spectra are in agreement with a model in which waves are driven unstable at the anomalous Doppler resonance, while only weakly damped at the Cerenkov resonance

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

Energy Technology Data Exchange (ETDEWEB)

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

1982-02-01

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

Toroidal Dielectric Tensor-Operator for Arbitrary Aspect-Ratio and Wave Frequency an Anisotropic-Resistivity MHD Formulation

International Nuclear Information System (INIS)

Komoshvili, K.; Cuperman, S.

1998-01-01

Motivated by the recently increased interest in small aspect ratio tokamaks, we have derived a 2(1/2)D dielectric tensor-operator which can properly describe the plasma response to r.f. waves, under conditions prevailing in the pre-heated stages of arbitrary aspect ratio, axisymmetric toroidal fusion devices. The derived dielectric tensor elements are based on a two-fluid, weakly collisional plasma description, with the Hall term included. They are characterized by the following features: (i) They are cast in a form evidencing the dielectric (non-operator) and operator contributions - the latter being due to the toroidal structure of the V-operators present in Maxwell's equations, on the background of equilibrium currents and pressure gradients; (ii) They are not subject to any I imitation on the (relative) magnitude of the toroidal effects - no expansion in the inverse aspect ratio parameter is used for their derivation; (iii) They include anisotropic - parallel and perpendicular to the magnetic field - contributions to the plasma resistivity; (iv) They are not Iimited by any restriction on the (relative) value of the wave frequency. The explicit, physically transparent formulation of the dielectric tensor is intended for the numerical solution of the full (E ll ≠ 0) wave equation and subsequently, evaluation of the Alfven wave current drive in small aspect ratio tokamaks

User's manual of Tokamak Simulation Code

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Nishino, Tooru; Tsunematsu, Toshihide; Sugihara, Masayoshi.

1992-12-01

User's manual for use of Tokamak Simulation Code (TSC), which simulates the time-evolutional process of deformable motion of axisymmetric toroidal plasma, is summarized. For the use at JAERI computer system, the TSC is linked with the data management system GAEA. This manual is forcused on the procedure for the input and output by using the GAEA system. Model equations to give axisymmetric motion, outline of code system, optimal method to get the well converged solution are also described. (author)

TBR-1 (Brazilian Tokamak) - Recent Results

International Nuclear Information System (INIS)

Fagundes, A.N.; Cruz Junior, D.F. da; Galvao, R.M.O.; Elizondo, J.I.; Nascimento, I.C. do; Sa, W.P. de; Sanada, E.K.; Silva, R.P.; Tuszel, A.G.; Vannucci, A.; Vuolo, J.H.

1987-08-01

The TBR-1 is a small Tokamak installed at the Physics Institute of the University of Sao Paulo. The machine was designed in 1977 and begun to be used in plasma scientific research in early 1980. its main characteristics are: Major radius, 0,30m; Minor radius (limiter), 0,08m; Toroidal field, 5 KG; Plasma current, 10KA (typical); Current duration, 6 ms (typical). In this paper we report the results of recent experimental research done in the TBR-1. (author) [pt

Expansion of parameter space for Toroidal Alfven Eigenmode experiments in TFTR

International Nuclear Information System (INIS)

Wong, K.L.; Wilson, J.R.; Chang, Z.Y.; Fredrickson, E.; Hammett, G.W.; Bush, C.; Nazikian, R.; Phillips, C.K.; Snipes, J.; Taylor, G.

1993-05-01

Several techniques were used to excite toroidal Alfven Eigenmodes in the Tokamak Fusion Test Reactor (TFTR) at magnetic fields above 10 kG. These involve pellet injection to raise the plasma density, variation of plasma current to change the energetic ion orbit and the q-profile, and ICRF heating to produce energetic hydrogen ions at velocities comparable to 3.5 MeV alpha particles. These experimental results are presented and relevance to fusion reactors are discussed

Stability of tearing modes in tokamak plasmas

International Nuclear Information System (INIS)

Hegna, C.C.; Callen, J.D.

1994-02-01

The stability properties of m ≥ 2 tearing instabilities in tokamak plasmas are analyzed. A boundary layer theory is used to find asymptotic solutions to the ideal external kink equation which are used to obtain a simple analytic expression for the tearing instability parameter Δ'. This calculation generalizes previous work on this topic by considering more general toroidal equilibria (however, toroidal coupling effects are ignored). Constructions of Δ' are obtained for plasmas with finite beta and for islands that have nonzero width. A simple heuristic estimate is given for the value of the saturated island width when the instability criterion is violated. A connection is made between the calculation of the asymptotic matching parameter in the finite beta and island width case to the nonlinear analog of the Glasser effect

Conceptual design of economic compact reversed shear Tokamak (CRST)

International Nuclear Information System (INIS)

Okano, Kunihiko

1998-01-01

Two indices of performance for economic analysis of Tokamak are defined as toroidal β value: β t (%)=(plasma pressure)/(pressure of magnetic field) and Troyon coefficient β N . The pressure of magnetic field is defined as β t 2 /2μ 0 (Bt: strength of toroidal magnetic field and μ 0 : permeability). β N is determined in order to make possible compare β t between other devices. To increase β N is very important on the economic viewpoint. ITER is designed as 2.2 β N , 1 MW/m 2 average neutron wall load, 8.14 m large radius and 2.8 m small radius, but the above values of CRST are 5.5, 4.5 MW/m 2 , 5.4 m and 1.59 m, respectively. Development of industrial and physical technologies makes possible to minimize economic Tokamak. After ITER, we expect that economic fusion reactor is obtained by minimization. CRST satisfies the conditions of economic fusion reactor conduced by the economic analysis. CRST is designed as 5.4 m main radius and 116x10 4 kW electric output. Fundamental physics and technologies, conceptual and industrial design of CRST are explained. (S.Y.)

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

Science.gov (United States)

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

2012-01-10

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

Fast-ion losses induced by ACs and TAEs in the ASDEX Upgrade tokamak

NARCIS (Netherlands)

M. García-Muñoz,; Hicks, N.; van Voornveld, R.; Classen, I.G.J.; Bilato, R.; Bobkov, V.; Brambilla, M.; Bruedgam, M.; Fahrbach, H. U.; Igochine, V.; Jaemsae, S.; Maraschek, M.; Sassenberg, K.

2010-01-01

The phase-space of convective and diffusive fast-ion losses induced by shear Alfven eigenmodes has been characterized in the ASDEX Upgrade tokamak. Time-resolved energy and pitch-angle measurements of fast-ion losses correlated in frequency and phase with toroidal Alfven eigenmodes (TAEs) and Alfven

Advanced transport modeling of toroidal plasmas with transport barriers

International Nuclear Information System (INIS)

Fukuyama, A.; Murakami, S.; Honda, M.; Izumi, Y.; Yagi, M.; Nakajima, N.; Nakamura, Y.; Ozeki, T.

2005-01-01

Transport modeling of toroidal plasmas is one of the most important issue to predict time evolution of burning plasmas and to develop control schemes in reactor plasmas. In order to describe the plasma rotation and rapid transition self-consistently, we have developed an advanced scheme of transport modeling based on dynamical transport equation and applied it to the analysis of transport barrier formation. First we propose a new transport model and examine its behavior by the use of conventional diffusive transport equation. This model includes the electrostatic toroidal ITG mode and the electromagnetic ballooning mode and successfully describes the formation of internal transport barriers. Then the dynamical transport equation is introduced to describe the plasma rotation and the radial electric field self-consistently. The formation of edge transport barriers is systematically studied and compared with experimental observations. The possibility of kinetic transport modeling in velocity space is also examined. Finally the modular structure of integrated modeling code for tokamaks and helical systems is discussed. (author)

Comparative studies of stellarator and tokamak transport

Energy Technology Data Exchange (ETDEWEB)

Stroth, U; Burhenn, R; Geiger, J; Giannone, L.; Hartfuss, H J; Kuehner, G; Ledl, L; Simmet, E E; Walter, H [Max-Planck-Inst. fuer Plasmaphysik, IPP-Euratom Association, Garching (Germany); ECRH Team; W7-AS Team

1997-09-01

Transport properties in the W7-AS stellarator and in tokamaks are compared. The parameter dependences and the absolute values of the energy confinement time are similar. Indications are found that the density dependence, which is usually observed in stellarator confinement, can vanish above a critical density. The density dependence in stellarators seems to be similar to that in the linear ohmic confinement regime, which, in small tokamaks, extends to high density values, too. Because of the similarity in the gross confinement properties, transport in stellarators and tokamaks should not be dominated by the parameters which are very different in the two concepts, i.e. magnetic shear, major rational values of the rotational transform and plasma current. A difference in confinement is that there exists evidence for pinches in the particle and, possibly, energy transport channels in tokamaks whereas in stellarators no pinches have been observed, so far. In order to study the effect of plasma current and toroidal electric fields, stellarator discharges were carried out with an increasing amount of plasma current. From these experiments, no clear evidence of a connection of pinches with these parameters is found. The transient response in W7-AS plasmas can be described in terms of a non-local model. As in tokamaks, also cold pulse experiments in W7-AS indicate the importance of non-local transport. (author). 8 refs, 5 figs.

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

International Nuclear Information System (INIS)

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

1986-08-01

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

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

NARCIS (Netherlands)

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

2009-01-01

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

Tokamak power system studies at ANL

International Nuclear Information System (INIS)

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

1986-06-01

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

Assembly study for JT-60SA tokamak

Energy Technology Data Exchange (ETDEWEB)

Shibanuma, K., E-mail: shibanuma.kiyoshi@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Arai, T.; Hasegawa, K.; Hoshi, R.; Kamiya, K.; Kawashima, H.; Kubo, H.; Masaki, K.; Saeki, H.; Sakurai, S.; Sakata, S.; Sakasai, A.; Sawai, H.; Shibama, Y.K.; Tsuchiya, K.; Tsukao, N.; Yagyu, J.; Yoshida, K.; Kamada, Y. [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Mizumaki, S. [Toshiba Corporation, Minato-ku, Tokyo 105-8001 (Japan); and others

2013-10-15

The assembly scenarios and assembly tools of the major tokamak components for JT-60SA are studied in the following. (1) The assembly frame (with a dedicated 30-tonne crane), which is located around the JT-60SA tokamak, is adopted for effective assembly works in the torus hall and the temporary support of the components during assembly. (2) Metrology for precise positioning of the components is also studied by defining the metrology points on the components. (3) The sector segmentation for weld joints and positioning of the vacuum vessel (VV), the assembly scenario and tools for VV thermal shield (TS), the connection of the outer intercoil structure (OIS) and the installation of the final toroidal field coil (TFC) are studied, as typical examples of the assembly scenarios and tools for JT-60SA.

Alpha-particle effects on high-n instabilities in tokamaks

International Nuclear Information System (INIS)

Rewoldt, G.

1988-06-01

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

Real time equilibrium reconstruction for tokamak discharge control

International Nuclear Information System (INIS)

Ferron, J.R.; Walker, M.L.; Lao, L.L.; St John, H.E.; Humphreys, D.A.; Leuer, J.A.

1998-01-01

A practical method for performing a tokamak equilibrium reconstruction in real time for arbitrary time varying discharge shapes and current profiles is described. An approximate solution to the Grad-Shafranov equilibrium relation is found which best fits the diagnostic measurements. Thus, a solution for the spatial distribution of poloidal flux and toroidal current density is available in real time that is consistent with plasma force balance, allowing accurate evaluation of parameters such as discharge shape and safety factor profile. The equilibrium solutions are produced at a rate sufficient for discharge control. This equilibrium reconstruction algorithm has been implemented on the digital plasma control system for the DIII-D tokamak. The first application of real time equilibrium reconstruction to discharge shape control is described. (author)

Numerical Calculation of Transport Based on the Drift-Kinetic Equation for Plasmas in General Toroidal Magnetic Geometry: Convergence and Testing; Calculo Numerico del Transporte mediante la Ecuacion Cinetica de Deriva para Plasmas en Geometria Magnetica Toroidal: Convergencia y Comprobaciones

Energy Technology Data Exchange (ETDEWEB)

Reynolds, J. M.; Lopez-Bruna, D.

2009-12-11

This report is the third of a series [Informes Tecnicos Ciemat 1165 y 1172] devoted to the development of a new numerical code to solve the guiding center equation for electrons and ions in toroidal plasmas. Two calculation meshes corresponding to axisymmetric tokamaks are now prepared and the kinetic equation is expanded so the standard terms of neoclassical theory --fi rst order terms in the Larmor radius expansion-- can be identified, restricting the calculations correspondingly. Using model density and temperature profiles for the plasma, several convergence test are performed depending on the calculation meshes and the expansions of the distribution function; then the results are compared with the theory [Hinton and Hazeltine, Rev. Mod. Phys. (1976)]. (Author) 18 refs.

Fast-ion losses induced by ACs and TAEs in the ASDEX Upgrade tokamak

NARCIS (Netherlands)

García-Munoz, M.; Hicks, N.; Voornveld, van R.; Classen, I.G.J.; Bilato, R.; Bobkov, V.; Brambilla, M.; Bruedgam, M.; Fahrbach, H. -U.; Igochine, V.; Jaemsae, S.; Maraschek, M.; Sassenberg, K.

2010-01-01

The phase-space of convective and diffusive fast-ion losses induced by shear Alfv´en eigenmodes has been characterized in the ASDEX Upgrade tokamak. Time-resolved energy and pitch-angle measurements of fast-ion losses correlated in frequency and phase with toroidal Alfv´en eigenmodes (TAEs) and

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

International Nuclear Information System (INIS)

Petrov, V.G.

1988-01-01

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

Reduction of toroidal magnetic field ripple in the advanced material tokamak experiment on JFT-2M

International Nuclear Information System (INIS)

Sato, M.; Miura, Y.; Kimura, H.; Yamamoto, M.; Koike, T.; Nakayama, T.; Hasegawa, M.; Urata, K.

1998-01-01

In order to reduce fast ion losses due to the toroidal field ripple, the reduction of ripple amplitude (δ) by inserting ferritic steel is studied, taking its toroidal mode number into account. The guideline of the design for reduction is wider and thicker ferritic board (FB) is located at further position from VV. The δ depends on the toroidal magnetic field. The value of B r21 /B t in the case of displacement of few cm is about 1 x 10 -5 which is one order smaller than the critical value. The offsetting of FB is not a problem for locked mode. Preliminary experiments with insertion of one or two FB's indicate no adverse effect on global plasma parameters. (author)

Reduction of toroidal magnetic field ripple in the advanced material tokamak experiment on JFT-2M

Energy Technology Data Exchange (ETDEWEB)

Sato, M.; Miura, Y.; Kimura, H.; Yamamoto, M.; Koike, T. [Japan Atomic Energy Research Inst. (Japan); Nakayama, T. [Hitachi Ltd. (Japan); Hasegawa, M. [Mitsubishi Electric Corp. (Japan); Urata, K. [Mitsubishi Heavy Industries Ltd. (Japan)

1998-07-01

In order to reduce fast ion losses due to the toroidal field ripple, the reduction of ripple amplitude ({delta}) by inserting ferritic steel is studied, taking its toroidal mode number into account. The guideline of the design for reduction is wider and thicker ferritic board (FB) is located at further position from VV. The {delta} depends on the toroidal magnetic field. The value of B{sub r21} /B{sub t} in the case of displacement of few cm is about 1 x 10{sup -5} which is one order smaller than the critical value. The offsetting of FB is not a problem for locked mode. Preliminary experiments with insertion of one or two FB's indicate no adverse effect on global plasma parameters. (author)

Experimental Evidence of Momentum Transport Induced by an Up-Down Asymmetric Magnetic Equilibrium in Toroidal Plasmas

International Nuclear Information System (INIS)

Camenen, Y.; Peeters, A. G.; Casson, F. J.; Hornsby, W. A.; Snodin, A. P.; Szepesi, G.; Bortolon, A.; Duval, B. P.; Federspiel, L.; Karpushov, A. N.; Piras, F.; Sauter, O.

2010-01-01

The first experimental evidence of parallel momentum transport generated by the up-down asymmetry of a toroidal plasma is reported. The experiments, conducted in the Tokamak a Configuration Variable, were motivated by the recent theoretical discovery of ion-scale turbulent momentum transport induced by an up-down asymmetry in the magnetic equilibrium. The toroidal rotation gradient is observed to depend on the asymmetry in the outer part of the plasma leading to a variation of the central rotation by a factor of 1.5-2. The direction of the effect and its magnitude are in agreement with theoretical predictions for the eight possible combinations of plasma asymmetry, current, and magnetic field.

Shear flow effects on ion thermal transport in tokamaks

International Nuclear Information System (INIS)

Tajima, T.; Horton, W.; Dong, J.Q.; Kishimoto, Y.

1995-03-01

From various laboratory and numerical experiments, there is clear evidence that under certain conditions the presence of sheared flows in a tokamak plasma can significantly reduce the ion thermal transport. In the presence of plasma fluctuations driven by the ion temperature gradient, the flows of energy and momentum parallel and perpendicular to the magnetic field are coupled with each other. This coupling manifests itself as significant off-diagonal coupling coefficients that give rise to new terms for anomalous transport. The authors derive from the gyrokinetic equation a set of velocity moment equations that describe the interaction among plasma turbulent fluctuations, the temperature gradient, the toroidal velocity shear, and the poloidal flow in a tokamak plasma. Four coupled equations for the amplitudes of the state variables radially extended over the transport region by toroidicity induced coupling are derived. The equations show bifurcations from the low confinement mode without sheared flows to high confinement mode with substantially reduced transport due to strong shear flows. Also discussed is the reduced version with three state variables. In the presence of sheared flows, the radially extended coupled toroidal modes driven by the ion temperature gradient disintegrate into smaller, less elongated vortices. Such a transition to smaller spatial correlation lengths changes the transport from Bohm-like to gyrobohm-like. The properties of these equations are analyzed. The conditions for the improved confined regime are obtained as a function of the momentum-energy deposition rates and profiles. The appearance of a transport barrier is a consequence of the present theory

Visible Spectrometer at the Compact Toroid Injection Experiment, the Sustained Spheromak Plasma Experiment and the Alcator C-Mod Tokamak for Doppler Width and Shift Measurements

Energy Technology Data Exchange (ETDEWEB)

Graf, A; Howard, S; Horton, R; Hwang, D; May, M; Beiersdorfer, P; McLean, H; Terry, J

2006-05-15

A novel Doppler spectrometer is currently being used for ion or neutral velocity and temperature measurements on the Alcator C-Mod Tokamak. The spectrometer has an f/No. of {approx}3.1 and is appropriate for visible light (3500-6700 {angstrom}). The full width at half maximum from a line emitting calibration source has been measured to be as small as 0.4 {angstrom}. The ultimate time resolution is line brightness light limited and on the order of ms. A new photon efficient detector is being used for the setup at C-Mod. Time resolution is achieved by moving the camera during a plasma discharge in a perpendicular direction through the dispersion plane of the spectrometer causing a vertical streaking across the camera face. Initial results from C-Mod as well as previous measurements from the Compact Toroid Injection Experiment (CTIX) and the Sustained Spheromak Plasma Experiment (SSPX) are presented.

Surface changes on probes inserted into the limiter shadow of the T-10 tokamak

International Nuclear Information System (INIS)

Chicherov, V.M.; Hildebrandt, D.; Laux, M.; Lingertat, J.; Lukianov, S.U.; Pech, P.; Reiner, H.D.; Stepanchikov, V.A.; Wolff, H.

1980-01-01

Time-resolved measurements of impurity fluxes which are oriented parallel and antiparallel to the toroidal and poloidal magnetic field have been performed in the limiter shadow of the T-10 tokamak. A qualitative model is proposed which explains the main features of the experimental results. (orig.)

Heat loads on poloidal and toroidal edges of castellated plasma-facing components in COMPASS

Science.gov (United States)

Dejarnac, R.; Corre, Y.; Vondracek, P.; Gaspar, J.; Gauthier, E.; Gunn, J. P.; Komm, M.; Gardarein, J.-L.; Horacek, J.; Hron, M.; Matejicek, J.; Pitts, R. A.; Panek, R.

2018-06-01

Dedicated experiments have been performed in the COMPASS tokamak to thoroughly study the power deposition processes occurring on poloidal and toroidal edges of castellated plasma-facing components in tokamaks during steady-state L-mode conditions. Surface temperatures measured by a high resolution infra-red camera are compared with reconstructed synthetic data from a 2D thermal model using heat flux profiles derived from both the optical approximation and 2D particle-in-cell (PIC) simulations. In the case of poloidal leading edges, when the contribution from local radiation is taken into account, the parallel heat flux deduced from unperturbed, upstream measurements is fully consistent with the observed temperature increase at the leading edges of various heights, respecting power balance assuming simple projection of the parallel flux density. Smoothing of the heat flux deposition profile due to finite ion Larmor radius predicted by the PIC simulations is found to be weak and the power deposition on misaligned poloidal edges is better described by the optical approximation. This is consistent with an electron-dominated regime associated with a non-ambipolar parallel current flow. In the case of toroidal gap edges, the different contributions of the total incoming flux along the gap have been observed experimentally for the first time. They confirm the results of recent numerical studies performed for ITER showing that in specific cases the heat deposition does not necessarily follow the optical approximation. Indeed, ions can spiral onto the magnetically shadowed toroidal edge. Particle-in-cell simulations emphasize again the role played by local non-ambipolarity in the deposition pattern.

Analysis of images from videocameras in the Frascati Tokamak Upgrade tokamak

International Nuclear Information System (INIS)

De Angelis, R.; Migliori, S.; Borioni, S.; Bracco, G.; Pierattini, S.; Perozziello, A.

2004-01-01

The plasma edge interaction in FTU tokamak is monitored by wide angle videocameras. Data are acquired as movies or single frames at a rate of 50 frames/s. The images show interesting features of the plasma such as the presence of Marfes or runaways and give useful information on the status of large parts of the vacuum vessel and toroidal limiter. Due to the large number of data available visual inspection of the movies is often insufficient to correlate the images to the experimental findings. This article illustrates a number of applications developed in order to correlate the images with plasma signals and to search the image database for specific features relevant to the discharge

Tokamak fluidlike equations, with applications to turbulence and transport in H mode discharges

International Nuclear Information System (INIS)

Kim, Y.B.; Biglari, H.; Carreras, B.A.; Diamond, P.H.; Groebner, R.J.; Kwon, O.J.; Spong, D.A.; Callen, J.D.; Chang, Z.; Hollenberg, J.B.; Sundaram, A.K.; Terry, P.W.; Wang, J.F.

1990-01-01

Significant progress has been made in developing tokamak fluidlike equations which are valid in all collisionality regimes in toroidal devices, and their applications to turbulence and transport in tokamaks. The areas highlighted in this paper include: the rigorous derivation of tokamak fluidlike equations via a generalized Chapman-Enskog procedure in various collisionality regimes and on various time scales; their application to collisionless and collisional drift wave models in a sheared slab geometry; applications to neoclassical drift wave turbulence; i.e. neoclassical ion-temperature-gradient-driven turbulence and neoclassical electron-drift-wave turbulence; applications to neoclassical bootstrap-current-driven turbulence; numerical simulation of nonlinear bootstrap-current-driven turbulence and tearing mode turbulence; transport in Hot-Ion H mode discharges. 20 refs., 3 figs

Time evolution of tokamak states with flow

International Nuclear Information System (INIS)

Kerner, W.; Weitzner, H.

1985-12-01

The general dissipative Braginskii single-fluid model is applied to simulate tokamak transport. An expansion with respect to epsilon = (ω/sub i/tau/sub i/) -1 , the factor by which perpendicular and parallel transport coefficients differ, yields a numerically tractable scheme. The resulting 1-1/2 D procedure requires computation of 2D toroidal equilibria with flow together with the solution of a system of ordinary 1D flux-averaged equations for the time evolution of the profiles. 13 refs

Microwave produced plasma in a Toroidal Device

Science.gov (United States)

Singh, A. K.; Edwards, W. F.; Held, E. D.

2010-11-01

A currentless toroidal plasma device exhibits a large range of interesting basic plasma physics phenomena. Such a device is not in equilibrium in a strict magneto hydrodynamic sense. There are many sources of free energy in the form of gradients in plasma density, temperature, the background magnetic field and the curvature of the magnetic field. These free energy sources excite waves and instabilities which have been the focus of studies in several devices in last two decades. A full understanding of these simple plasmas is far from complete. At Utah State University we have recently designed and installed a microwave plasma generation system on a small tokamak borrowed from the University of Saskatchewan, Saskatoon, Canada. Microwaves are generated at 2.45 GHz in a pulsed dc mode using a magnetron from a commercial kitchen microwave oven. The device is equipped with horizontal and vertical magnetic fields and a transformer to impose a toroidal electric field for current drive. Plasmas can be obtained over a wide range of pressure with and without magnetic fields. We present some preliminary measurements of plasma density and potential profiles. Measurements of plasma temperature at different operating conditions are also presented.

Nonideal magnetohydrodynamic instabilities and toroidal magnetic confinement

International Nuclear Information System (INIS)

Furth, H.P.

1985-05-01

The marked divergence of experimentally observed plasma instability phenomena from the predictions of ideal magnetohydrodynamics led in the early 1960s to the formulations of finite-resistivity stability theory. Beginning in the 1970s, advanced plasma diagnostics have served to establish a detailed correspondence between the predictions of the finite-resistivity theory and experimental plasma behavior - particularly in the case of the resistive kink mode and the tokamak plasma. Nonlinear resistive-kink phenomena have been found to govern the transport of magnetic flux and plasma energy in the reversed-field pinch. The other predicted finite-resistivity instability modes have been more difficult to identify directly and their implications for toroidal magnetic confinement are still unresolved

Progress in toroidal confinement and fusion research

International Nuclear Information System (INIS)

Furth, H.P.

1987-10-01

During the past 30 years, the characteristic T/sub i/n tau/sub E/-value of toroidal-confinement experiments has advanced by more than seven orders of magnitude. Part of this advance has been due to an increase of gross machine parameters. Most of this advance has been due to an increase of gross machine parameters. Most of the advance is associated with improvements in the ''quality of plasma confinement.'' The combined evidence of spherator and tokamak research clarifies the role of magnetic-field geometry in determining confinement and points to the importance of shielding out plasma edge effects. A true physical understanding of anomalous transport remains to be achieved. 39 refs., 11 figs., 1 tab

Eigenvalues of relaxed toroidal plasmas of arbitrary sharp edged cross sections. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Khalil, Sh M [Plasma Physics and Nuclear Fusion Department, Nuclear Research Center, Atomic Energy Authority, Cairo, (Egypt)

1996-03-01

Relaxed (force-free) toroidal plasmas described by the equations cur 1 B={mu}B, and grad {mu}=O (B is the magnetic field) induces interest in nuclear fusion. Its solution is perceived to describe the gross of the reversed field pinch (RFP), spheromak configuration, current limitation in toroidal plasmas, and others. The parameter {mu} plays an important roll in relaxed states. It cannot exceed the smallest eigenvalue {mu} (min), and that for a toroidal discharge there is a maximum toroidal current which is connected to this value. The values of{mu} were calculated numerically, using the methods of collocation points, for toroids of arbitrary aspect ratio {alpha} ({alpha} = R/a, ratio of major/minor radii of tokamak) and arbitrary curved cross-sections (circle, ellipse, cassini, and D-shaped). The aim of present work is to prove the applicability of the numerical methods for calculating the eigenvalues for toroidal plasmas having sharp edged cross sections and arbitrary aspect ratio. The lowest eigenvalue {mu} (min), satisfy the boundary condition {beta} .n = O (or RB. = O) for which the toroidal flux are calculated. These are the zero field eigenvalues of the equation cur 1 b={mu}B. The poloidal magnetic field lines corresponding to different cross sections are shown by plotting the boundary condition B.n=O. The plots showed good fulfillment of the boundary condition along the whole boundaries of different cross sections. Dependence of eigenvalues {mu}a on aspect ratio {alpha} is also obtained. Several runs of the programme with various wave numbers K showed that {mu}a is very insensitive to the choice of K. 8 figs.

Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

Science.gov (United States)

Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

2018-05-01

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

Atomic physics effects on dissipative toroidal drift wave stability

International Nuclear Information System (INIS)

Beer, M.A.; Hahm, T.S.

1992-02-01

The effects of atomic physics processes such as ionization, charge exchange, and radiation on the linear stability of dissipative drift waves are investigated in toroidal geometry both numerically and analytically. For typical TFTR and TEXT edge parameters, overall linear stability is determined by the competition between the destabilizing influence of ionization and the stabilizing effect due to the electron temperature gradient. An analytical expression for the linear marginal stability condition, η e crit , is derived. The instability is most likely to occur at the extreme edge of tokamaks with a significant ionization source and a steep electron density gradient

Threshold for the destabilisation of the ion-temperature-gradient mode in magnetically confined toroidal plasmas

Science.gov (United States)

Zocco, A.; Xanthopoulos, P.; Doerk, H.; Connor, J. W.; Helander, P.

2018-02-01

The threshold for the resonant destabilisation of ion-temperature-gradient (ITG) driven instabilities that render the modes ubiquitous in both tokamaks and stellarators is investigated. We discover remarkably similar results for both confinement concepts if care is taken in the analysis of the effect of the global shear . We revisit, analytically and by means of gyrokinetic simulations, accepted tokamak results and discover inadequacies of some aspects of their theoretical interpretation. In particular, for standard tokamak configurations, we find that global shear effects on the critical gradient cannot be attributed to the wave-particle resonance destabilising mechanism of Hahm & Tang (Phys. Plasmas, vol. 1, 1989, pp. 1185-1192), but are consistent with a stabilising contribution predicted by Biglari et al. (Phys. Plasmas, vol. 1, 1989, pp. 109-118). Extensive analytical and numerical investigations show that virtually no previous tokamak theoretical predictions capture the temperature dependence of the mode frequency at marginality, thus leading to incorrect instability thresholds. In the asymptotic limit , where is the rotational transform, and such a threshold should be solely determined by the resonant toroidal branch of the ITG mode, we discover a family of unstable solutions below the previously known threshold of instability. This is true for a tokamak case described by a local local equilibrium, and for the stellarator Wendelstein 7-X, where these unstable solutions are present even for configurations with a small trapped-particle population. We conjecture they are of the Floquet type and derive their properties from the Fourier analysis of toroidal drift modes of Connor & Taylor (Phys. Fluids, vol. 30, 1987, pp. 3180-3185), and to Hill's theory of the motion of the lunar perigee (Acta Math., vol. 8, 1886, pp. 1-36). The temperature dependence of the newly determined threshold is given for both confinement concepts. In the first case, the new temperature

Discharge initiation experiments in the Tokapole II tokamak

International Nuclear Information System (INIS)

Shepard, D.A.

1984-06-01

Experiments in the Tokapole II tokamak demonstrate the benefits of high density (n/sub e//n 0 greater than or equal to 0.01) preionization by reducing four quantities at startup: necessary toroidal loop voltage (V 1 ) (50%), volt-second consumption (40 to 50%), impurity radiation (25 to 50%), and runaway electron production (approx. 80 to 100%). A zero-dimensional code models the loop voltage reduction dependence on preionization density and predicts a similar result for reactor scale devices. The code shows low initial resistivity and a high resistivity time derivative contribute to loop voltage reduction. The power balance of the ECR plasma in a toroidal-field-only case was studied. Langmuir probes and impurity doping were used. The vertical electric field (E/sub v/) and current (I/sub v/), which result from curvature drift, were measured (E/sub v/ approx. 10 V/cm and I/sub v/ approx. 50 Amps) and exceeded expected values for the bulk electron temperature (approx. 10 eV). A series of experiments with external windings to simulate field errors perpendicular to the toroidal field was done. The results imply that an error field of 0.1% of the toroidal field is deleterious to ECR plasma density

Pulse discharge cleaning of the vacuum vessel of HL-1 tokamak

International Nuclear Information System (INIS)

Li Guodong; Zhu Yukun; Xiao Zhenggui; Sun Shouqi; Ze Mingrui

1986-01-01

The HL-1 Tokamak was test-operated on September 21, 1984. During the period of vacuum conditioning, including 60 hours of baking up to 200 deg C and 7 x 10 4 shots of pulse discharge cleaning, the calculated quantities of carbon and oxygen removed are equivalent to 24 and 6 monolayers, respectively. Then, 124 shots of tokamak discharge were performed with low level plasma parameters. The plasma current and pulse length achieved were 60 kA and 85 ms at the toroidal magnetic field of 15 kG. This paper described the techniques used and the effect on discharge characteristics of bakeout and pulse discharge cleaning of the vacuum vessel

Report on the high magnetic field tokamak TRIAM-1

Energy Technology Data Exchange (ETDEWEB)

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

1981-02-01

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

Existence of core localized toroidicity-induced Alfven eigenmode

International Nuclear Information System (INIS)

Fu, G.Y.

1995-02-01

The core-localized toroidicity-induced Alfven eigenmode (TAE) is shown to exist at finite plasma pressure due to finite aspect ratio effects in tokamak plasma. The new critical beta for the existence of the TAE mode is given by α∼ 3ε + 2s 2 , where ε = r/R is the inverse aspect ratio, s is the magnetic shear and α = -Rq 2 dβ/dr is the normalized pressure gradient. In contrast, previous critical α is given by α ∼ s 2 . In the limit of s << √r/R, the new critical α is greatly enhanced by the finite aspect ratio effects

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

International Nuclear Information System (INIS)

Ivers, T.H.

1991-01-01

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

Fast-ion transport induced by Alfvén eigenmodes in the ASDEX Upgrade tokamak

DEFF Research Database (Denmark)

Garcia-Munoz, M.; Classen, I.G.J.; Geiger, B.

2011-01-01

A comprehensive suite of diagnostics has allowed detailed measurements of the Alfvén eigenmode (AE) spatial structure and subsequent fast-ion transport in the ASDEX Upgrade (AUG) tokamak [1]. Reversed shear Alfvén eigenmodes (RSAEs) and toroidal induced Alfvén eigenmodes (TAEs) have been driven u...

Dynamics of spheromak-like compact toroids in a drift tube

International Nuclear Information System (INIS)

Suzuki, Y.; Kishimoto, Y.; Hayashi, T.

2001-01-01

In order to supply plasma fuel confined in spheromak-like compact toroids (SCTs) to a fusion device, the SCTs must be successfully guided through a drift tube region, in which they might be influenced by the magnetic field leaking from the fusion device. To reveal the SCT dynamics in a drift tube, MHD numerical simulations, where the SCTs are accelerated in a co-axial perfectly conducting cylinder with an external magnetic field, are carried out. In addition, the effect of an extended central electrode is examined by changing the length of the inner conducting cylinder. It is revealed that the SCT penetration depth is shorter than that estimated from the conventional conducting sphere model and that the SCTs are further decelerated by extending the inner conducting cylinder. These results are consistent with the results of the compact toroid injection experiment performed on the TEXT Upgrade tokamak. Finally, the deceleration mechanism of the SCTs is discussed by comparing the simulation result with the proposed theoretical model. (author)

New results from Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.K.

2002-01-01

New results from Globus-M spherical tokamak (ST) are presented. Reported are the achievements of high plasma current of 0.36 MA and high toroidal magnetic field of 0.55 T. Plasma column stability in Globus-M is conserved at low edge safety factors and high plasma densities. Achieved lowest safety factor was q(cyl) 19 m -3 . New methods of density increase are discussed. Low-density boarder of operational space is investigated. Runaway electrons properties and conditions of their generation are investigated. Results look promising for STs. Plasma-wall interaction study was performed. Silicon probes were installed into vacuum vessel. They were exposed to boronization, first, and then deposited film interacted with plasma. Discussed are film properties. Briefly described are new diagnostic tools installed on tokamak. Status and preliminary results obtained with auxiliary heating systems are shown. (author)

Turbulence and abnormal transport in tokamak plasmas

International Nuclear Information System (INIS)

Garbet, X.

1988-06-01

The objective of this thesis is the study of plasma microinstabilities in linear and nonlinear tokamak regime. After a brief review of experimental results the theoretical tools used in this study are presented. A variational method founded on the existence of angular variables system and on action for charged particles in tokamak configurations is detailed. The correspondent functional extreme with regard to fluctuating electromagnetic field, is calculated analytically with taking into account the toroidal geometry. A numerical code, TORRID, has been constructed on this principle and the main instabilities, particularly ionic instabilities and microtearing, has been linearly studied. The most simple non linear methods are rewieved and applied at the microtearing instabilities. The quasilinear transport coefficients are deducted of an entropy minimum production principle. The ionic thermic conductivity and the viscosity are calculated for an ionic turbulence [fr

Resonant helical fields in the TBR tokamak

International Nuclear Information System (INIS)

Bender, O.W.

1986-01-01

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

Axisymmetric stability of vertically asymmetric Tokamaks at large beta poloidal

Energy Technology Data Exchange (ETDEWEB)

Yamazaki, K.; Fishman, H.; Okabayashi, M. (Princeton Univ., NJ (USA). Plasma Physics Lab.); Todd, A.M.M. (Grumman Aerospace Corp., Princeton, NJ (USA))

1983-11-01

The rigid-mode stability of high-..beta.. vertically asymmetric Tokamak equilibria with quasi-uniform current profile is investigated analytically using toroidicity-shaping double expansion method. It is found that vertical stability at large beta poloidal is mainly determined by a coupling between the shape of the plasma surface and the Shafranov shift of the magnetic axis. To the lowest order, symmetric components of the plasma surface shape are found to be the critical destabilizing elements. Asymmetric components have little effect. The inclusion of higher order terms in the high-..beta.. Tokamak expansion leads to further destabilization. These analytic insights are qualitatively confirmed by numerical stability calculations using the PEST code with parabolic safety-factor profile.

Experimental study of toroidicity-induced Alfven eigenmode (TAE) stability at high q(0)

International Nuclear Information System (INIS)

Batha, S.H.; Levinton, F.M.; Spong, D.A.

1995-07-01

Experiments to destabilize the Toroidicity-induced Alfven Eigenmode (TAE) by energetic alpha particles were performed on the Tokamak Fusion Test Reactor using deuterium and tritium fuel. To decrease the alpha particle pressure instability threshold, discharges with an elevated value of q(0) > 1.5 were used. By raising q(0), the radial location of the low toroidal-mode-number TAE gaps moves toward the magnetic axis and into alignment with the region of maximum alpha pressure gradient, thereby (in theory) lowering the value of β α (0) required for instability. No TAE activity was observed when the central alpha particle β α reached 0.08% in a discharge with fusion power of 2.4 MW. Calculations show that the fusion power is within a factor of 1.5 to 3 of the instability threshold

Merging startup experiments on the UTST spherical tokamak

International Nuclear Information System (INIS)

Yamada, Takuma; Kamio, Shuji; Imazawa, Ryota

2010-01-01

The University of Tokyo Spherical Tokamak (UTST) was constructed to explore the formation of ultrahigh-beta spherical tokamak (ST) plasmas using double null plasma merging. The main feature of the UTST is that the poloidal field coils are located outside the vacuum vessel to demonstrate startup in a reactor-relevant situation. Initial operations used partially completed power supplies to investigate the appropriate conditions for plasma merging. The plasma current of the merged ST reached 100 kA when the central solenoid coil was used to assist plasma formation. Merging of two ST plasmas through magnetic reconnection was successfully observed using two-dimensional pickup coil arrays, which directly measure the toroidal and axial magnetic fields inside the UTST vacuum vessel. The resistivity of the current sheet was found to be anomalously high during merging. (author)

Preliminary results of MHD stability in HL-1 tokamak

International Nuclear Information System (INIS)

Zheng Yongzhen; Ma Tengcai; Xiao Zhenggui Cai Renfang

1987-01-01

In this paper, MHD activities of HL-1 tokamak plasma are studied with Fourier transform and correlatio analysis. The poloidal modes m = 1, 2, 3,4 and toroidal modes n of MHD magnetic fluctuation signals are detected. Methods for suppressing MHD instabilities are suggested and tested, after MHD instabilities are studied in HL-1. The effects of MHD characteristics in the beginning stage of discharge on the whole process of discharge are analyzed. The disruption, in HL-1 device could be divided into three kinds: internal disruption, minor disruption and major disruption. The result shows that HL-1 will have a better operation condition if internal disruption appears. In is end, the stable operation region of HL-1 tokamak is also given

High-field, high-density tokamak power reactor

International Nuclear Information System (INIS)

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

1977-11-01

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

Integral equation based stability analysis of short wavelength drift modes in tokamaks

International Nuclear Information System (INIS)

Hirose, A.; Elia, M.

2003-01-01

Linear stability of electron skin-size drift modes in collisionless tokamak discharges has been investigated in terms of electromagnetic, kinetic integral equations in which neither ions nor electrons are assumed to be adiabatic. A slab-like ion temperature gradient mode persists in such a short wavelength regime. However, toroidicity has a strong stabilizing influence on this mode. In the electron branch, the toroidicity induced skin-size drift mode previously predicted in terms of local kinetic analysis has been recovered. The mode is driven by positive magnetic shear and strongly stabilized for negative shear. The corresponding mixing length anomalous thermal diffusivity exhibits favourable isotope dependence. (author)

Runaway electron generation in tokamak disruptions

International Nuclear Information System (INIS)

Helander, P.; Andersson, F.; Fueloep, T.; Smith, H.; Anderson, D.; Lisak, M.; Eriksson, L.-G.

2005-01-01

The time evolution of the plasma current during a tokamak disruption is calculated by solving the equations for runaway electron production simultaneously with the induction equation for the toroidal electric field. The resistive diffusion time in a post-disruption plasma is typically comparable to the runaway avalanche growth time. Accordingly, the toroidal electric field induced after the thermal quench of a disruption diffuses radially through the plasma at the same time as it accelerates runaway electrons, which in turn back-react on the electric field. When these processes are accounted for in a self-consistent way, it is found that (1) the efficiency and time scale of runaway generation agrees with JET experiments; (2) the runaway current profile typically becomes more peaked than the pre-disruption current profile; and (3) can easily become radially filamented. It is also shown that higher runaway electron generation is expected if the thermal quench is sufficiently fast. (author)

Three-dimensional features of GAM zonal flows in the HL-2A tokamak

International Nuclear Information System (INIS)

Yan, L.W.; Cheng, J.; Hong, W.Y.; Zhao, K.J.; Lan, T.; Dong, J.Q.; Liu, A.D.; Yu, C.X.; Yu, D.L.; Qian, J.; Huang, Y.; Yang, Q.W.; Ding, X.T.; Liu, Y.; Pan, C.H.

2007-01-01

A novel design of the three-step Langmuir probe (TSLP) array has been developed to investigate the zonal flow (ZF) physics in the HL-2A tokamak. Three TSLP arrays are applied to measure the three-dimensional (3D) features of ZFs. They are separated by 65 mm in the poloidal and 800 mm in the toroidal directions, respectively. The 3D properties of the geodesic acoustic mode (GAM) ZFs are presented. The poloidal and toroidal modes of the radial electric fields of the GAM perturbations are simultaneously determined in the HL-2A tokamak for the first time. The modes have narrow radial wave numbers (k r ρ i = 0.03-0.07) and short radial scale lengths (2.4-4.2 cm). High coherence of both the GAM and the ambient turbulence separated by toroidal 22.5 0 along a magnetic field line is observed, which contrasts with the high coherence of the GAM and the low coherence of the ambient turbulence apart from the field line. The nonlinear three wave coupling between the turbulent fluctuations and the ZFs is a plausible mechanism for flow generation. The skewness and kurtosis spectra of the probability distribution function of the potential perturbations are contrasted with the corresponding bicoherence for the first time, which support the three wave coupling mechanism

Generación y dinámica de electrones runaway en plasmas tokamak

OpenAIRE

Fernández Gómez, Isabel

2016-01-01

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

Wavelength calibration of x-ray imaging crystal spectrometer on Joint Texas Experimental Tokamak

International Nuclear Information System (INIS)

Yan, W.; Chen, Z. Y.; Jin, W.; Huang, D. W.; Ding, Y. H.; Li, J. C.; Zhang, X. Q.; Zhuang, G.; Lee, S. G.; Shi, Y. J.

2014-01-01

The wavelength calibration of x-ray imaging crystal spectrometer is a key issue for the measurements of plasma rotation. For the lack of available standard radiation source near 3.95 Å and there is no other diagnostics to measure the core rotation for inter-calibration, an indirect method by using tokamak plasma itself has been applied on joint Texas experimental tokamak. It is found that the core toroidal rotation velocity is not zero during locked mode phase. This is consistent with the observation of small oscillations on soft x-ray signals and electron cyclotron emission during locked-mode phase

Strength-limited magnetic field intensity of toroidal magnet systems fabricated or the base of layer-by-layer shrouded solenoids

International Nuclear Information System (INIS)

Litvinnko, Yu.A.

1982-01-01

The possibilities, as to the ultimate magnetic field strength, of tokamak magnet systems made on the base of layer-by-laeyer shrouded coils are considered numerically. The toroidal magnet system is considered which consists of N skewe, layer-by-layer shrouded, equistrong coils in the ideal torus approximation. The dependences of the ragnetic field strength on the internal- and external torus radii, pulse duration and aspect ratio for copper coils shrouded with fiberglass are calculated as an example. The analysis of the obtained results shows that using of the layer-by-layer shrouding scheme for toroidal solenoid coils leads to a considerable growth of the ultimate magnetic field strengths in a wide duration range. For example, the limiting field strength along the toroidal solenoid axis of the considered type inside the ''FT'' installation toroidal solenoid at equivalent field pulse duration of approximately 0.3 s reaches H 0 =1.3zx10 7 A/m

A New Interpretation of Alpha-particle-driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

R. Nazikian; G.J. Kramer; C.Z. Cheng; N.N. Gorelenkov; H.L. Berk; S.E. Sharapov

2003-01-01

The original description of alpha-particle-driven instabilities in the Tokamak Fusion Test Reactor (TFTR) in terms of Toroidal Alfvin Eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the anti-ballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time

Control strategy for plasma equilibrium in a tokamak

International Nuclear Information System (INIS)

Miskell, R.V.

1975-01-01

The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)

MHD stability limits in the TCV Tokamak

International Nuclear Information System (INIS)

Reimerdes, H.

2001-07-01

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 a > 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

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

Neutronics calculations for the Oak Ridge National Laboratory Tokamak Reactor Studies

International Nuclear Information System (INIS)

Santoro, R.T.; Baker, V.C.; Barnes, J.M.

1976-01-01

Neutronics calculations have been carried out to analyze the nuclear performance of conceptual blanket and shield designs for the Tokamak Experimental Power Reactor (EPR) and the Tokamak Demonstration Reactor Plant (DRP) being considered at the Oak Ridge National Laboratory. These reactor designs represent a sequence in the commercialization of fusion-generated electrical power. All of the calculations were carried out using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV coupled neutron-gamma-ray transport cross-section data, fluence-to-kerma conversion factors, and radiation damage cross-section data. The calculations include spatial and integral heating-rate estimates in the reactor with emphasis on the recovery of fusion neutron energy in the blanket and limiting the heat-deposition rate in the superconducting toroidal field coils. Radiation damage due to atomic displacements and gas production produced in the reactor structural material and in the toroidal field coil windings were also estimated. The tritium-breeding ratio when natural lithium is used as the fertile material in the DRP blanket and in the experimental breeding modules in the EPR is also given

The superconducting magnet system for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Lang, D.D.; Bulmer, R.J.; Chaplin, M.R.; O'Connor, T.G.; Slack, D.S.; Wong, R.L.; Zbasnik, J.P.; Schultz, J.H.; Diatchenko, N.; Montgomery, D.B.

1994-01-01

The superconducting magnet system for the Tokamak Physics eXperiment (TPX) will be the first all superconducting magnet system for a Tokamak, where the poloidal field coils, in addition to the toroidal field coils are superconducting. The magnet system is designed to operate in a steady state mode, and to initiate the plasma discharge ohmically. The toroidal field system provides a peak field of 4.0 Tesla on the plasma axis at a plasma major radius of 2.25 m. The peak field on the niobium 3-tin, cable-in-conduit (CIC) conductor is 8.4 Tesla for the 16 toroidal field coils. The toroidal field coils must absorb approximately 5 kW due to nuclear heating, eddy currents, and other sources. The poloidal field system provides a total of 18 volt seconds to initiate the plasma and drive a plasma current up to 2 MA. The poloidal field system consists of 14 individual coils which are arranged symmetrically above and below the horizontal mid plane. Four pairs of coils make up the central solenoid, and three pairs of poloidal ring coils complete the system. The poloidal field coils all use a cable-in-conduit conductor, using either niobium 3-tin (Nb 3 Sn) or niobium titanium (NbTi) superconducting strands depending on the operating conditions for that coil. All of the coils are cooled by flowing supercritical helium, with inlet and outlet connections made on each double pancake. The superconducting magnet system has gone through a conceptual design review, and is in preliminary design started by the LLNL/MIT/PPPL collaboration. A number of changes have been made in the design since the conceptual design review, and are described in this paper. The majority of the design and all fabrication of the superconducting magnet system will be ,accomplished by industry, which will shortly be taking over the preliminary design. The magnet system is expected to be completed in early 2000

Neoclassical toroidal viscosity in perturbed equilibria with general tokamak geometry

Energy Technology Data Exchange (ETDEWEB)

Logan, Nikolas C.; Park, Jong-Kyu; Kim, Kimin; Wang, Zhirui [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Berkery, John W. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

2013-12-15

This paper presents a calculation of neoclassical toroidal viscous torque independent of large-aspect-ratio expansions across kinetic regimes. The Perturbed Equilibrium Nonambipolar Transport (PENT) code was developed for this purpose, and is compared to previous combined regime models as well as regime specific limits and a drift kinetic δf guiding center code. It is shown that retaining general expressions, without circular large-aspect-ratio or other orbit approximations, can be important at experimentally relevant aspect ratio and shaping. The superbanana plateau, a kinetic resonance effect recently recognized for its relevance to ITER, is recovered by the PENT calculations and shown to require highly accurate treatment of geometric effects.

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

International Nuclear Information System (INIS)

Reid, R.L.

1983-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-05-15

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

Effects of magnetic shear on current penetration in a tokamak

International Nuclear Information System (INIS)

Zhang Pengyun; Wang Long

2001-01-01

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

Remote servicing considerations for near term tokamak power reactors (TNS). Final summary

International Nuclear Information System (INIS)

Spampinato, P.T.

1977-01-01

Next generation Tokamaks require special consideration for remote servicing. Three major problems are highlighted: (1) movement of heavy components, (2) remote connection/disconnection of joints, and (3) remote cutting, welding, and leak detection. The first problem is assumed to be handled with existing expertise and is not considered. The remaining problems are thought to be minimized by considering two engineering departures from conventional tokamak design; locating the field shaping coils outside of the toroidal coils and enclosing the total device within an evacuated reactor cell. Five topics under this vacuum building concept are discussed: incremental cost, vacuum pumping, tritium containment, activation topology, and first year operations

Transport and dynamics in toroidal fusion systems. Report of second year progress, 1993--1994

International Nuclear Information System (INIS)

Schnack, D.D.

1994-01-01

In this document the author describes an extension of the spatial gridding techniques to an MHD model suitable for the description of the dynamics of toroidal fusion devices. Since the dominant MHD modes in these devices have relatively long toroidal wavelength, the toroidal coordinate is approximated with finite Fourier series. The unstructured, triangular mesh is used to describe the details of the poloidal geometry. With some exceptions, the hydrodynamic variables are treated in a manner analogous to that used in CFD. These quantities (mass, energy, and momentum) are volume based densities that satisfy scalar or vector conservation laws. The electromagnetic variables (the magnetic flux density B and the electric current density J) are area based densities that satisfy pseudo-vector conservation laws, and have no counterpart in fluid dynamics. These variables are also constrained to remain solenoidal. These quantities are represented on the triangular mesh in a new manner that is an extension of that used on rectangular, structured meshes. In this work the author has chosen to solve the primitive MHD equations in order to make the resulting codes and techniques more generally applicable to problems beyond the narrow scope of tokamak plasmas. The temporal stiffness problems inherent in this description of tokamak dynamics that motivate the reduced MHD model are addressed here with the semi-implicit method of time integration. Finally, the author remarks that, while the present work deals strictly with the MHD equations, other volume based fluid descriptions, such as diffusive transport could easily be adapted to these techniques and coupled with the description of the electromagnetic field presented here

Stability of the Global Alfven Eigenmode in the presence of fusion alpha particles in an ignited tokamak plasma

International Nuclear Information System (INIS)

Fu, G.Y.; Van Dam, J.W.

1989-05-01

The stability of the Global Alfven Eigenmodes is investigated in the presence of super-Alfvenic energetic particles, such as the fusion-product alpha particles in an ignited deuterium-tritium tokamak plasma. Alpha particles tend to destabilize these modes when ω *α > ω A , where ω A is the shear-Alfven modal frequency and ω *α is the alpha particle diamagnetic drift frequency. This destabilization due to alpha particles is found to be significantly enhanced when the alpha particles are modeled with a slowing-down distribution function rather than with a Maxwellian. However, previously neglected electron damping due to the magnetic curvature drift is found to be comparable in magnitude to the destabilizing alpha particle term. Furthermore, the effects of toroidicity are also found to be stabilizing, since the intrinsic toroidicity induces poloidal mode coupling, which enhances the parallel electron damping from the sideband shear-Alfven Landau resonance. In particular, for the parameters of the proposed Compact Ignition Tokamak, the Global Alfven Eigenmodes are found to be completely stabilized by either the electron damping that enters through the magnetic curvature drift or the damping introduced by finite toroidicity. 29 refs., 8 figs., 1 tab

Sawtooth-induced loss of runaway electrons in tokamaks

International Nuclear Information System (INIS)

Yan Longwen; Shi Bingren; Jiao Yiming

2001-01-01

A model based on banana orbit loss has been proposed to explain the sawtooth effect on the loss of the runaway electrons in tokamaks. Circulating runaway electrons can be transferred into the trapped ones due to magnetic perturbation during sawtooth crashes, then they are repelled to the limiter via toroidal precession drift with a time delay. This model may also clarify the hard X-ray oscillations correlated with the m = 2 mode and the hard X-ray bursts during outer disruptions

Non-Solenoidal Startup Research Directions on the Pegasus Toroidal Experiment

Science.gov (United States)

Fonck, R. J.; Bongard, M. W.; Lewicki, B. T.; Reusch, J. A.; Winz, G. R.

2017-10-01

The Pegasus research program has been focused on developing a physical understanding and predictive models for non-solenoidal tokamak plasma startup using Local Helicity Injection (LHI). LHI employs strong localized electron currents injected along magnetic field lines in the plasma edge that relax through magnetic turbulence to form a tokamak-like plasma. Pending approval, the Pegasus program will address a broader, more comprehensive examination of non-solenoidal tokamak startup techniques. New capabilities may include: increasing the toroidal field to 0.6 T to support critical scaling tests to near-NSTX-U field levels; deploying internal plasma diagnostics; installing a coaxial helicity injection (CHI) capability in the upper divertor region; and deploying a modest (200-400 kW) electron cyclotron RF capability. These efforts will address scaling of relevant physics to higher BT, separate and comparative studies of helicity injection techniques, efficiency of handoff to consequent current sustainment techniques, and the use of ECH to synergistically improve the target plasma for consequent bootstrap and neutral beam current drive sustainment. This has an ultimate goal of validating techniques to produce a 1 MA target plasma in NSTX-U and beyond. Work supported by US DOE Grant DE-FG02-96ER54375.

Effects of Density and Impurity on Edge Localized Modes in Tokamaks

Science.gov (United States)

Zhu, Ping

2017-10-01

Plasma density and impurity concentration are believed to be two of the key elements governing the edge tokamak plasma conditions. Optimal levels of plasma density and impurity concentration in the edge region have been searched for in order to achieve the desired fusion gain and divertor heat/particle load mitigation. However, how plasma density or impurity would affect the edge pedestal stability may have not been well known. Our recent MHD theory modeling and simulations using the NIMROD code have found novel effects of density and impurity on the dynamics of edge-localized modes (ELMs) in tokamaks. First, previous MHD analyses often predict merely a weak stabilizing effect of toroidal flow on ELMs in experimentally relevant regimes. We find that the stabilizing effects on the high- n ELMs from toroidal flow can be significantly enhanced with the increased edge plasma density. Here n denotes the toroidal mode number. Second, the stabilizing effects of the enhanced edge resistivity due to lithium-conditioning on the low- n ELMs in the high confinement (H-mode) discharges in NSTX have been identified. Linear stability analysis of the experimentally constrained equilibrium suggests that the change in the equilibrium plasma density and pressure profiles alone due to lithium-conditioning may not be sufficient for a complete suppression of the low- n ELMs. The enhanced resistivity due to the increased effective electric charge number Zeff after lithium-conditioning provides additional stabilization of the low- n ELMs. These new effects revealed in our theory analyses may help further understand recent ELM experiments and suggest new control schemes for ELM suppression and mitigation in future experiments. They may also pose additional constraints on the optimal levels of plasma density and impurity concentration in the edge region for H-mode tokamak operation. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB

Magnetohydrodynamic Stability of a Toroidal Plasma's Separatrix

International Nuclear Information System (INIS)

Webster, A. J.; Gimblett, C. G.

2009-01-01

Large tokamaks capable of fusion power production such as ITER, should avoid large edge localized modes (ELMs), thought to be triggered by an ideal magnetohydrodynamic instability due to current at the plasma's separatrix boundary. Unlike analytical work in a cylindrical approximation, numerical work finds the modes are stable. The plasma's separatrix might stabilize modes, but makes analytical and numerical work difficult. We generalize a cylindrical model to toroidal separatrix geometry, finding one parameter Δ ' determines stability. The conformal transformation method is generalized to allow nonzero derivatives of a function on a boundary, and calculation of the equilibrium vacuum field allows Δ ' to be found analytically. As a boundary more closely approximates a separatrix, we find the energy principle indicates instability, but the growth rate asymptotes to zero

Energy confinement of tokamak plasma with consideration of bootstrap current effect

International Nuclear Information System (INIS)

Yuan Ying; Gao Qingdi

1992-01-01

Based on the η i -mode induced anomalous transport model of Lee et al., the energy confinement of tokamak plasmas with auxiliary heating is investigated with consideration of bootstrap current effect. The results indicate that energy confinement time increases with plasma current and tokamak major radius, and decreases with heating power, toroidal field and minor radius. This is in reasonable agreement with the Kaye-Goldston empirical scaling law. Bootstrap current always leads to an improvement of energy confinement and the contraction of inversion radius. When γ, the ratio between bootstrap current and total plasma current, is small, the part of energy confinement time contributed from bootstrap current will be about γ/2

Development path of low aspect ratio tokamak power plants

International Nuclear Information System (INIS)

Stambaugh, R.D.; Chan, V.S.; Miller, R.L.

1997-03-01

Recent advances in tokamak physics indicate the 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% beta toroidal with the normalized beta as high as 10 and fully aligned 100% bootstrap current. Such beta 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 (Q PLANT ) of 1.9. The high beta values in the ST mean that E x B shear stabilization of turbulence should be 10 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 beta 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 Q PLANT 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. The favorable size scaling of the ST and the high beta mean that in large sizes, if the copper TF coil is replaced with a superconducting TF coil and a shield, the advanced fuel D-He 3 could be burned in a device with Q PLANT ∼ 4

Effect of halo current and its toroidal asymmetry during disruptions in JT-60U

International Nuclear Information System (INIS)

Neyatani, Y.; Yoshino, R.; Ando, T.

1995-01-01

A poloidal halo current due to a vertical displacement event (VDE) is observed in experimentally simulated VDE discharges and density limit disruptions in the JT-60U tokamak. In the case of a clockwise I p and B T discharge, the halo current flows into the vacuum vessel from the inside separatrix and goes back to the plasma from the outside separatrix. A maximum halo current is produced by a change in the poloidal flux generated by plasma current decay. A toroidal asymmetry factor of 2.5 is estimated from the requirements of the fracture of the carbon-fiber composite tiles. The toroidal asymmetry is caused by the poloidal field (PF) that is produced by the toroidal field (TF) ripple, the deformation of the vacuum vessel, the setting error between the vacuum vessel and the TF and PF coils, the low-n mode during current quench, etc. To consider this asymmetry, in JT-60U, one must estimate the total halo current as nearly 26% of the plasma current just before a current quench. 25 refs., 10 figs

A versatile ray-tracing code for studying rf wave propagation in toroidal magnetized plasmas

International Nuclear Information System (INIS)

Peysson, Y; Decker, J; Morini, L

2012-01-01

A new ray-tracing code named C3PO has been developed to study the propagation of arbitrary electromagnetic radio-frequency (rf) waves in magnetized toroidal plasmas. Its structure is designed for maximum flexibility regarding the choice of coordinate system and dielectric model. The versatility of this code makes it particularly suitable for integrated modeling systems. Using a coordinate system that reflects the nested structure of magnetic flux surfaces in tokamaks, fast and accurate calculations inside the plasma separatrix can be performed using analytical derivatives of a spline-Fourier interpolation of the axisymmetric toroidal MHD equilibrium. Applications to reverse field pinch magnetic configuration are also included. The effects of 3D perturbations of the axisymmetric toroidal MHD equilibrium, due to the discreteness of the magnetic coil system or plasma fluctuations in an original quasi-optical approach, are also studied. Using a Runge–Kutta–Fehlberg method for solving the set of ordinary differential equations, the ray-tracing code is extensively benchmarked against analytical models and other codes for lower hybrid and electron cyclotron waves. (paper)

Toroidal rotation braking with n = 1 magnetic perturbation field on JET

International Nuclear Information System (INIS)

Sun, Y; Liang, Y; Koslowski, H R; Harting, D; Wiegmann, C; Wiesen, S; Jachmich, S; Alfier, A; Asunta, O; Corrigan, G; Giroud, C; Gryaznevich, M P; Hender, T; Nardon, E; Parail, V; Naulin, V; Tala, T

2010-01-01

A strong toroidal rotation braking has been observed in plasmas with application of an n = 1 magnetic perturbation field on the JET tokamak. Calculation results from the momentum transport analysis show that the torque induced by the n = 1 perturbation field has a global profile. The maximal value of this torque is at the plasma core region (ρ - √ν regime in the plasma core, but it is close to the transition between the 1/ν and ν - √ν regimes. The neoclassical toroidal viscosity (NTV) torque in the 1/ν and ν - √ν regimes is calculated. The observed torque is of a magnitude in between that of the NTV torque in the 1/ν and ν - √ν regimes. The NTV torque in the ν - √ν regimes is enhanced using the Lagrangian variation of the magnetic field strength. However, it is still smaller than the observed torque by one order of magnitude.

Tokamak experimental power reactor

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

1978-01-01

A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

Stochastization of Magnetic Field Surfaces in Tokamaks by an Inner Coil

International Nuclear Information System (INIS)

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

2006-01-01

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

Importance of Plasma Response to Non-axisymmetric Perturbations in Tokamaks

International Nuclear Information System (INIS)

Park, Jong-kyu; Boozer, Allen H.; Menard, Jonathan E.; Garofalo, Andrea M.; Schaffer, Michael J.; Hawryluk, Richard J.; Kaye, Stanley M.; Gerhardt, Stefan P.; Sabbagh, Steve A. and the NSTX Team

2009-01-01

Tokamaks are sensitive to deviations from axisymmetry as small as (delta)B/B 0 ∼ 10 -4 . These non-axisymmetric perturbations greatly modify plasma confinement and performance by either destroying magnetic surfaces with subsequent locking or deforming magnetic surfaces with associated non-ambipolar transport. The Ideal Perturbed Equilibrium Code (IPEC) calculates ideal perturbed equilibria and provides important basis for understanding the sensitivity of tokamak plasmas to perturbations. IPEC calculations indicate that the ideal plasma response, or equivalently the effect by ideally perturbed plasma currents, is essential to explain locking experiments on National Spherical Torus eXperiment (NSTX) and DIII-D. The ideal plasma response is also important for Neoclassical Toroidal Viscosity (NTV) in non-ambipolar transport. The consistency between NTV theory and magnetic braking experiments on NSTX and DIII-D can be improved when the variation in the field strength in IPEC is coupled with generalized NTV theory. These plasma response effects will be compared with the previous vacuum superpositions to illustrate the importance. However, plasma response based on ideal perturbed equilibria is still not sufficiently accurate to predict the details of NTV transport, and can be inconsistent when currents associated with a toroidal torque become comparable to ideal perturbed currents

The recent research progress on the J-TEXT tokamak

International Nuclear Information System (INIS)

Wang, Z.J.; Zhuang, G.; Gentle, K.W.

2013-01-01

The recent research progress on the J-TEXT tokamak is introduced. The interaction between resonant magnetic perturbations (RMPs) and plasma have been carried out on the J-TEXT tokamak and the results show that the m/n = 2/1 (m and n are the poloidal and toroidal mode numbers, respectively) mode locking is obtained with sufficiently large RMPs while suppression of the m/n = 2/1 tearing mode by moderate magnetic perturbation amplitude is also observed. With a model based on reduced magnetohydrodynamics (MHD) equations, both the mode locking and mode suppression by RMPs are simulated and the results are in good agreement with the experimental observations. To observe the current profile, a high resolution three-wave far infrared polarimeter/interferometer is set up and the first results indicate it works well. (author)

On the density limit of Tokamaks

International Nuclear Information System (INIS)

Lehnert, B.

1982-12-01

Under the conditions of so far performed quasi-steady tokamak experiments near the density limit, the plasma pressure gradient in the outer layers of the plasma body becomes mainly determined by the plasma-neutral gas balance. An earlier analysis of ballooning instabilities driven by this gradient in regions of bad curvature has been extended to deduce an explicit stability criterion which determines the density limit. This criterion is closely related to the empirical Murakami limit. At relevant tokamak data, the deduced limit becomes proportional to J(sub)zR(sup)1/2 where J(sub)z is the average current density and R the major plasma radius. It is further found to be independent of the toroidal magnetic field strength and anomalous transport, as well as to be a slow function of the outer layer temperature and the mass number. The deduced stability criterion is consistent with so far performed experiments. Provided that the present analysis can be extrapolated to a wider range of parameter data and be combined with Alcator scaling, conditions near ignition appear to become realizable in small tokamaks by ohmic heating alone. These conditions can be satisfied at relevant magnetic field strengths and plasma currents, by imposing a high plasma current density. (author)

Toroidal field magnet and poloidal divertor field coil systems adapted to reactor requirements

International Nuclear Information System (INIS)

Koeppendoerfer, W.

1985-01-01

ASDEX Upgrade is a tokamak experiment with external poloidal field coils, that is now under construction at IPP Garching. It can produce elongated single-null (SN), double-null (DN) and limiter (L) configurations. The SN is the reference configuration with asymmetric load distributions in the poloidal field (PF) system and the toroidal field (TF) magnet. Plasma control and stabilization requires a rigid passive conductor close to the plasma. The design principles of the coils and support structure are described. (orig.)

Theory of nonaxisymmetric vertical displacement events in tokamaks

International Nuclear Information System (INIS)

Fitzpatrick, R.

2011-01-01

A semi-analytic sharp-boundary model of a nonaxisymmetric vertical displacement event (VDE) in a large aspect-ratio, high-beta (i.e. β ∼ ε), vertically elongated tokamak plasma is developed. The model is used to simulate nonaxisymmetric VDEs with a wide range of different plasma equilibrium and vacuum vessel parameters. These simulations yield poloidal halo current fractions and toroidal peaking factors whose magnitudes are similar to those seen in experiments, and also reproduce the characteristic inverse scaling between the halo current fraction and the toroidal peaking factor. Moreover, the peak poloidal halo current density in the vacuum vessel is found to correlate strongly with the reciprocal of the minimum edge safety factor attained during the VDE. In addition, under certain circumstances, the ratio of the net sideways force acting on the vacuum vessel to the net vertical force is observed to approach unity. Finally, the peak vertical force per unit area acting on the vessel is found to have a strong correlation with the equilibrium toroidal plasma current at the start of the VDE, but is also found to increase with increasing vacuum vessel resistivity relative to the scrape-off layer plasma.

Topics in stability and transport in tokamaks: Dynamic transition to second stability with auxiliary heating; stability of global Alfven waves in an ignited plasma

International Nuclear Information System (INIS)

Fu, G.

1988-01-01

The problem of access to the high-beta ballooning second-stability regime by means of auxiliary heating and the problem of the stability of global-shear Alfven waves in an ignited tokamak plasma are theoretically investigated. These two problems are related to the confinement of both the bulk plasma as well as the fusion-product alpha particles an dare fundamentally important to the operation of ignited tokamak plasma. First, a model that incorporates both transport and ballooning mode stability was developed in order to estimate the auxiliary heating power required for tokamak plasma to evolve in time self-consistently into a high-beta, globally self-stabilized equilibrium. The critical heating power needed for access to second stability is found to be proportional to the square root of the anomalous diffusivity induced by the ballooning instability. Next, the full effects of toroidicity are retained in a theoretical description of global-type-shear Alfven modes whose stability can be modified by the fusion-product alpha particles that will present in an ignited tokamak plasma. Toroidicity is found to induce mode coupling and to stabilize the so-called Global Alfven Eigenmodes (GAE)

Applicability of the PHITS code to a tokamak fusion device

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko; Okuno, Koichi; Kawasaki, Hiromitsu

2011-01-01

The three-dimensional Monte-Carlo code PHITS (particle and Heavy Ion Transport code System) has been developed to perform the radiation transport analysis, design of the radiation shields and neutronics calculations for tokamak-type D-D fusion reactors. A subroutine was included in PHITS to represent the toroidal neutron source of 2.45 MeV neutrons from the D-D reaction. Here, an example of preliminary tests using PHITS is given. (author)

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

International Nuclear Information System (INIS)

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

1975-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Edge plasma physical investigations of tokamak plasmas in CRIP

International Nuclear Information System (INIS)

Bakos, J.; Ignacz, P.; Koltai, L.; Paszti, F.; Petravich, G.; Szigeti, J.; Zoletnik, S.

1988-01-01

The results of the measurements performed in the field of thermonuclear high temperature plasma physics in CRIP (Hungary) are summarized. In the field of the edge plasma physics solid probes were used to test the external zone of plasma edges, and atom beams and balls were used to investigate both the external and internal zones. The plasma density distribution was measured by laser blow-off technics, using Na atoms, which are evaporated by laser pulses. The excitation of Na atom ball by tokamak plasma gives information on the status of the plasma edge. The toroidal asymmetry of particle transport in tokamak plasma was measured by erosion probes. The evaporated and transported impurities were collected on an other part of the plasma edge and were analyzed by SIMS and Rutherford backscattering. The interactions in plasma near the limiter were investigated by a special limiter with implemented probes. Recycling and charge exchange processes were measured. Disruption phenomena of tokamak plasma were analyzed and a special kind of disruptions, 'soft disruptions' and the related preliminary perturbations were discovered. (D.Gy.) 10 figs

ITER tokamak device

International Nuclear Information System (INIS)

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

1991-01-01

The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

Equilibrium and stability of high-beta toroidal plasmas with toroidal and poloidal flow in reduced magnetohydrodynamic models

International Nuclear Information System (INIS)

Ito, A.; Nakajima, N.

2010-11-01

Effects of flow, finite ion temperature and pressure anisotropy on equilibrium and stability of a high-beta toroidal plasma are studied in the framework of reduced magnetohydrodynamics (MHD). A set of reduced equilibrium equations for high-beta tokamaks with toroidal and poloidal flow comparable to the poloidal sound velocity is derived in a unified form of single-fluid and Hall MHD models and a two-fluid MHD model with ion finite Larmor radius (FLR) terms. Pressure anisotropy is introduced with equations for the parallel heat flux which are closed by a fluid closure model. It is solved analytically for the single-fluid model and the solutions shows complicated characteristics in the region around the poloidal sound velocity due to pressure anisotropy and the parallel heat flux. Numerical solutions are found by using the finite element method for the two-fluid model with FLR effects in the case of isotropic, adiabatic pressure and indicate the following features of two-fluid equilibria: the isosurfaces of the magnetic flux, the pressure and the ion stream function do not coincide with each other, and the solutions depend on the sign of the radial electric field. Reduced single-fluid MHD equations with time evolution that are consistent with the above equilibria are also derived in order to study their stability. They conserve the energy up to the order required by the equilibria. (author)

Influence of a minor periodicity on the magnetic island formation in tokamaks

International Nuclear Information System (INIS)

Matsuda, Shinzaburo

1975-01-01

A formation of magnetic islands due to external error fields in tokamaks is described. In particular, current control fields associated with shell gaps are shown to make islands of significant size. Moreover, we found that a toroidal minor periodicity of these perturbation fields, which is generally represented by the number of shell gaps, has an important meaning for the suppression of the resonant magnetic islands. (auth.)

Effects of density asymmetries on heavy-impurity transport in a rotating tokamak-plasma

International Nuclear Information System (INIS)

Romanelli, M.; Ottaviani, M.

1997-12-01

The transport equations of heavy trace-impurities in a Tokamak plasma with strong toroidal rotation have been studied analytically in the collisional regime. It is found that the poloidal asymmetry of the impurity-density, which occurs because of the rotation, brings about a large enhancement of the diffusivity and indeed of the pinch velocity above the conventional Pfirsh-Schlueter values. (author)

Review of the Advanced Toroidal Facility program

International Nuclear Information System (INIS)

Lyon, J.F.; Murakami, M.

1987-01-01

This report summarizes the history and design goals of the Advanced Toroidal Facility (ATF). The ATF is nearing completion at ORNL with device completion expected in May 1987 and first useful plasma operation in June/July 1987. ATF is a moderate-aspect-ratio torsatron, the world's largest stellarator facility with R = 2.1 m, α bar = 0.3 m and B = 2 T (5-s pulse) or 1 T (steady-state capability). It has been specifically designed to support the US tokamak program by studying important toroidal confinement issues in a similar magnetic geometry that allows external control of the magnetic configuration properties and their radial profiles: transform, shear, well depth, shaping, axis topology, etc. ATF will operate in a current-free model which allows separation of current-driven and pressure-driven plasma behavior. It also complements the world stellarator program in its magnetic configuration (between Heliotron-E and W VII-AS) and its capabilities (large size, good access, steady state capability, second stability access, etc.). For both roles ATF will require high-power long-pulse heating to carry out its physics goals since the high power NBI pulse is limited to 0.3 s. The ATF program focuses on demonstrating the principles of high-beta, steady-state operation in toroidal geometry through its study of: (1) scaling of beta limits with magnetic configuration properties and the plasma behavior in the second stability regime; (2) transport scaling at low collisionality and the role/control of electric field; (3) control of plasma density and impurities using divertors; (4) plasma heating with NBI, ECH, ICH, and plasma fueling with gas puffing and pellet injection; and (5) optimization of the magnetic configuration

Designing a Sine-Coil for Measurement of Plasma Displacements in IR-T1 Tokamak

International Nuclear Information System (INIS)

Khorshid, Pejman; Razavi, M.; Molaii, M.; Ghoranneviss, M.; TalebiTaher, A.; Arvin, R.; Mohammadi, S.; NikMohammadi, A.

2008-01-01

A method for the measurement of the plasma position in the IR-T1 tokamak in toroidal coordinates is developed. A sine-coil, which is a Rogowski coil with a variable wiring density is designed and fabricated for this purpose. An analytic solution of the Biot-Savart law, which is used to calculate magnetic fields created by toroidal plasma current, is presented. Results of calculations are compared with the experimental data obtained in no-plasma shots with a toroidal current-carrying coil positioned inside the vessel to simulate the plasma movements. The results are shown a good linear behavior of plasma position measurements. The error is less than 2.5% and it is compared with other methods of measurements of the plasma position. This method will be used in the feedback position control system and tests of feedback controller parameters are ongoing

Magnetic fluctuations associated with density fluctuations in the tokamak edge

International Nuclear Information System (INIS)

Kim, Y.J.; Gentle, K.W.; Ritz, C.P.; Rhodes, T.L.; Bengtson, R.D.

1989-01-01

Electrostatic density and potential fluctuations occurring with high amplitude near the edge of a tokamak are correlated with components of the fluctuating magnetic field measured outside the limiter radius. It has been established that this turbulence is associated with fluctuations in current as well as density and potential. The correlation extends for substantial toroidal distances, but only if the probes are displaced approximately along field lines, consistent with the short coherence lengths poloidally but long coherence lengths parallel to the field which are characteristic for this turbulence. Furthermore, the correlation can be found only with density fluctuations measured inside the limiter radius; density fluctuations behind the limiter have no detectable magnetic concomitant for the toroidally spaced probes used here. (author). Letter-to-the-editor. 12 refs, 3 figs

A theoretical study of impurity production at limiters in tokamaks

International Nuclear Information System (INIS)

Pitcher, C.S.; Matthews, G.F.; Goodall, D.H.J.; McCracken, G.M.; Stangeby, P.C.

1988-01-01

The spatial distribution of neutral impurity emissions around graphite limiters in the DITE tokamak is investigated using a Monte Carlo neutral transport code based on physical sputtering. The Monte Carlo code results are compared with experiments for the CI distributions observed toroidally around the probe limiter and radially around the fixed limiter. The comparison between code and experiment demonstrates the ability of camera observations and Langmuir probe measurements to provide detailed information on the production of impurities at limiters. From the Monte Carlo simulation it is shown that the toroidal distribution of impurity emission is determined mainly by the geometry of the limiter and the radial variation of the sputtering yield, while the radial distribution is sensitive mainly to the sputtered atom velocity distribution and the rate coefficients for ionization and excitation

Device for supporting a toroidal coil in a toroidal type nuclear fusion device

International Nuclear Information System (INIS)

Kitazawa, Hakaru; Sato, Hiroshi.

1975-01-01

Object: To easily manufacture a center block having a strength sufficient to withstand an electromagnetic force exerted on the center of toroidal of a toroidal coil and to increase its reliability. Structure: In a device for supporting toroidal coils wherein the electromagnetic force exerted on the center of toroidal of a plurality of toroidal coils arranged in toroidal fashion, the contact surface between the toroidal coil and the center block is arranged parallel to the center axis of toroidal so as to receive the electromagnetic force exerted on the center of toroidal of the toroidal coil as the component of force in a radial direction. (Taniai, N.)

Toroidal fusion reactor design based on the reversed-field pinch

International Nuclear Information System (INIS)

Hagenson, R.L.

1978-07-01

The toroidal reversed-field pinch (RFP) achieves gross equilibrium and stability with a combination of high shear and wall stabilization, rather than the imposition of tokamak-like q-constraints. Consequently, confinement is provided primarily by poloidal magnetic fields, poloidal betas as large as approximately 0.58 are obtainable, the high ohmic-heating (toroidal) current densities promise a sole means of heating a D-T plasma to ignition, and the plasma aspect ratio is not limited by stability/equilibrium constraints. A reactor-like plasma model has been developed in order to quantify and to assess the general features of a power system based upon RFP confinement. An ''operating point'' has been generated on the basis of this plasma model and a relatively detailed engineering energy balance. These results are used to generate a conceptual engineering model of the reversed-field pinch reactor (RFPR) which includes a general description of a 750 MWe power plant and the preliminary consideration of vacuum/fueling, first wall, blanket, magnet coils, iron core, and the energy storage/transfer system

Stress analyses of ITER toroidal field coils under fault conditions

International Nuclear Information System (INIS)

Jong, C.T.J.

1990-02-01

The International Thermonuclear Experimental Reactor (ITER) is intended as an experimental thermonuclear tokamak reactor for testing the basic physics, performance and technologies essential to future fusion reactors. The ITER design will be based on extensive new design work, supported by new physical and technological results, and on the great body of experience built up over several years from previous national and international reactor studies. Conversely, the ITER design process should provide the fusion community with valuable insights into what key areas need further development or clarification as we move forward towards practical fusion power. As part of the design process of the ITER toroidal field coils the mechanical behaviour of the magnetic system under fault conditions has to be analysed in more detail. This paper describes the work carried out to create a detailed finite element model of two toroidal field coils as well as some results of linear elastic analyses with fault conditions. The analyses have been performed with the finite element code ANSYS. (author). 5 refs.; 8 figs.; 2 tabs

Plasma confinement of Nagoya high-beta toroidal-pinch experiments

International Nuclear Information System (INIS)

Hirano, K.; Kitagawa, S.; Wakatani, M.; Kita, Y.; Yamada, S.; Yamaguchi, S.; Sato, K.; Aizawa, T.; Osanai, Y.; Noda, N.

1977-01-01

Two different types of high-β toroidal pinch experiments, STP [1] and CCT [2,3], have been done to study the confinement of the plasma produced by a theta-pinch. The STP is an axisymmetric toroidal pinch of high-β tokamak type, while the CCT consists of multiply connected periodic toroidal traps. Internal current-carrying copper rings are essential to the CCT. Since both apparatuses use the same fast capacitor bank system, they produce rather similar plasma temperatures and densities. The observed laser scattering temperature and density is about 50 eV and 4x10 15 cm -3 , respectively, when the filling pressure is 5 mtorr. In the STP experiment, strong correlations are found between the βsub(p) value and the amplitude of m=2 mode. It has a minimum around the value of βsub(p) of 0.8. The disruptive instability is observed to expand the pinched plasma column without lowering the plasma temperature. Just before the disruption begins, the q value around the magnetic axis becomes far less than 1 and an increase of the amplitude of m=2 mode is seen. The CCT also shows rapid plasma expansion just before the magnetic field reaches its maximum. Then the trap is filled up with the plasma by this irreversible expansion and stable plasma confinement is achieved. The energy confinement time of the CCT is found to be about 35 μs. (author)

Simulation of runaway electron generation and diffusion during major disruptions in the HL-2A tokamak

International Nuclear Information System (INIS)

Li, Yanli; Sun, Jizhong; Zhang, Yipo; Sang, Chaofeng; Wu, Na; Wang, Dezhen

2014-01-01

Highlights: • The strong and long duration magnetic perturbation (δB/B ∼ 1.0 × 10 −3 ) can restrain the RE generation effectively. • The REs are generated initially in the plasma core during disruptions. • The toroidal electric field does not exhibit a centrally hollow phenomenon. • The toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field. - Abstract: The generation and diffusion of runaway electrons (REs) during major disruptions in the HL-2A tokamak has been studied numerically. The diffusion caused by the magnetic perturbation is especially addressed. The simulation results show that the strong magnetic perturbation (δB/B ∼ 1.0 × 10 −3 ) can cause a significant loss of REs due to the radial diffusion and restrain the RE avalanche effectively. The results also indicate that the REs are generated initially in the plasma core during disruptions, and that the toroidal electric field does not exhibit a centrally hollow phenomenon. In addition, it is found that the toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field

Cross-section sensitivity analyses for a Tokamak Experimental Power Reactor

International Nuclear Information System (INIS)

Simmons, E.L.; Gerstl, S.A.W.; Dudziak, D.J.

1977-09-01

The objectives of this report were (1) to determine the sensitivity of neutronic responses in the preliminary design of the Tokamak Experimental Power Reactor by Argonne National Laboratory, and (2) to develop the use of a neutron-gamma coupled cross-section set in the calculation of cross-section sensitivity analysis. Response functions such as neutron plus gamma kerma, Mylar dose, copper transmutation, copper dpa, and activation of the toroidal field coil dewar were investigated. Calculations revealed that the responses were most sensitive to the high-energy group cross sections of iron in the innermost regions containing stainless steel. For example, both the neutron heating of the toroidal field coil and the activation of the toroidal field coil dewar show an integral sensitivity of about -5 with respect to the iron total cross sections. Major contributors are the scattering cross sections of iron, with -2.7 and -4.4 for neutron heating and activation, respectively. The effects of changes in gamma cross sections were generally an order of 10 lower

Disruptions in Tokamaks

International Nuclear Information System (INIS)

Bondeson, A.

1987-01-01

This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)

Cryogenic analysis of forced-cooled, superconducting TF magnets for compact tokamak reactors

International Nuclear Information System (INIS)

Kerns, J.A.; Slack, D.S.; Miller, J.R.

1988-01-01

Current designs for compact tokamak reactors require the toroidal- field (TF) superconducting magnets to produce fields from 10 to 15 T at the winding pack, using high-current densities to high nuclear heat loads (greater than 1 kW/coil in some instances), which are significantly greater than the conduction and radiation heat loads for which cryogenic systems are usually designed. A cryogenic system for the TF winding pack for two such tokamak designs has been verified by performing a detailed, steady-state heat-removal analysis. Helium properties along the forced-cooled conductor flow path for a range of nuclear heat loads have been calculated. The results and implications of this analysis are presented. 12 refs., 6 figs

Toroidally asymmetric ELM precursor oscillations in the TCV tokamak

International Nuclear Information System (INIS)

Reimerdes, H.; Pochelon, A.; Guittienne, P.; Weisen, H.; Suttrop, W.

1997-01-01

In TCV ohmic H-modes have been obtained in diverted single-null (SND), double-null (DND), and elongated limited plasma configurations. In ELM-free H-modes the particle density rises continuously until the discharge usually terminates with a high density disruption. Quasi-stationary H-modes have been obtained in the presence of ELMs. The observed ELM spectrum is continuous and ranges from clearly identifiable type III ELMs to low frequency, large ELMs. The necessity of ELMs for particle control of H-mode plasmas while causing high peak-power loads on strike points makes the control of their level and nature desirable and motivates the study of the underlying MHD-instability. Prior to ELMs in TCV coherent magnetic oscillations, that indicate a rapidly growing MHD instability, have been observed. The structure of these precursor oscillation is investigated with TCV's Mirnov probe arrays. In particular an observed toroidal asymmetry in the growth of the instability has to be explained. (author) 2 figs., 6 refs

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

International Nuclear Information System (INIS)

CHU, M.S; PARKS, P.B

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Transport through dissipative trapped electron mode and toroidal ion temperature gradient mode in TEXTOR

International Nuclear Information System (INIS)

Rogister, A.; Hasselberg, G.; Waelbroeck, F.; Weiland, J.

1987-12-01

A self-consistent transport code is used to evaluate how plasma confinement in tokamaks is influenced by the microturbulent fields which are excited by the dissipative trapped electron (DTE) instability. As shown previously, the saturation theory on which the code is based has been developed from first principles. The toroidal coupling resulting from the ion magnetic drifts is neglected; arguments are presented to justify this approximation. The numerical results reproduce well the neo-Alcator scaling law observed experimentally - e.g. in TEXTOR - in non detached ohmic discharges, the confinement degradation which results when auxiliary heating is applied, as well as a large number of other experimental observations. We also assess the possible impact of the toroidal ion temperature gradient mode on energy confinement by estimating the ion thermal flux with the help of the mixing length approximation. (orig./GG)

New dual gas puff imaging system with up-down symmetry on experimental advanced superconducting tokamak

DEFF Research Database (Denmark)

Liu, S. C.; Shao, L. M.; Zweben, S. J.

2012-01-01

advanced superconducting tokamak (EAST). The two views are up-down symmetric about the midplane and separated by a toroidal angle of 66.6 degrees. A linear manifold with 16 holes apart by 10 mm is used to form helium gas cloud at the 130x130 mm (radial versus poloidal) objective plane. A fast camera...

Propagation and damping of mode converted ion-Bernstein waves in toroidal plasmas

International Nuclear Information System (INIS)

Ram, A.K.; Bers, A.

1991-01-01

In the heating of tokamak plasmas by waves in the ion-cyclotron range of frequencies, the fast Alfven waves launched at the plasma edge can mode convert to the ion-Bernstein waves (IBW). The propagation and damping of these mode converted waves was studied using a ray tracing code that follows the fast phase and the amplitude of the electromagnetic field along the IBW ray trajectories in a toroidal plasma. A simple analytical model is developed that describes the numerically observed features of propagation and damping of the IBW's. It is found that along the ray trajectory of the IBW there is an upshift of the poloidal mode numbers, which can lead to the electron Landau damping of the wave. This damping is dependent on the strength of the toroidal plasma current. From the properties of the upshift of the poloidal mode numbers, it is concluded that the mode converted ion-Bernstein waves are not suitable candidates for electron current drive

Observation of finite-β MHD phenomena in Tokamaks

International Nuclear Information System (INIS)

McGuire, K.M.

1985-01-01

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 qsub(o) 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

Design considerations for ITER toroidal field coils

International Nuclear Information System (INIS)

Kalsi, S.S.; Lousteau, D.C.; Miller, J.R.

1987-01-01

The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Europe, Japan, the Union of Soviet Socialist Republics (U.S.S.R.), and the United States. This paper describes a magnetic and mechanical design methodology for toroidal field (TF) coils that employs Nb 3 Sn superconductor technology. Coil winding is sized by using conductor concepts developed for the U.S. TIBER concept. Manifold concepts are presented for the complete cooling system. Also included are concepts for the coil structural arrangement. The effects of in-plane and out-of-plane loads are included in the design considerations for the windings and case. Concepts are presented for reacting these loads with a minimum amount of additional structural material. Concepts discussed in this paper could be considered for the ITER TF coils

The ARIES tokamak fusion reactor study

International Nuclear Information System (INIS)

Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

1989-01-01

The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

Ohmic ignition of Neo-Alcator tokamak with adiabatic compression

International Nuclear Information System (INIS)

Inoue, Nobuyuki; Ogawa, Yuichi

1992-01-01

Ohmic ignition condition on axis of the DT tokamak plasma heated by minor radius and major radius adiabatic compression is studied assuming parabolic profiles for plasma parameters, elliptic plasma cross section, and Neo-Alcator confinement scaling. It is noticeable that magnetic compression reduces the necessary total plasma current for Ohmic ignition device. Typically in compact ignition tokamak of the minor radius of 0.47 m, major radius of 1.5 m and on-axis toroidal field of 20 T, the plasma current of 6.8 MA is sufficient for compression plasma, while that of 11.7 MA is for no compression plasma. Another example with larger major radius is also described. In such a device the large flux swing of Ohmic transformer is available for long burn. Application of magnetic compression saves the flux swing and thereby extends the burn time. (author)

Overview, Progress, and Plans for the Compact Toroidal Hybrid Experiment

Science.gov (United States)

Hartwell, G. J.; Allen, N. R.; Ennis, D. A.; Hanson, J. D.; Howell, E. C.; Johnson, C. A.; Knowlton, S. F.; Kring, J. D.; Ma, X.; Maurer, D. A.; Ross, K. G.; Schmitt, J. C.; Traverso, P. J.; Williamson, E. N.

2017-10-01

The Compact Toroidal Hybrid (CTH) is an l = 2 , m = 5 torsatron/tokamak hybrid (R0 = 0.75 m, ap 0.2 m, and | B | disruption studies. The main goals of the CTH experiment are to study disruptive behavior as a function of applied 3D magnetic shaping, and to test and advance the V3FIT reconstruction code and NIMROD modeling of CTH. The disruptive density limit is observed to exceed the Greenwald limit as the vacuum transform is increased with no observed threshold for avoidance. Low-q operations (1.1 routine, with disruptions ceasing if the vacuum transform is raised above 0.07. Sawteeth are observed in CTH and have a similar phenomenology to tokamak sawteeth despite employing a 3D confining field. Application of vacuum transform has been demonstrated to reduce and eliminate the vertical drift of elongated discharges. Internal SXR diagnostics, in conjunction with external magnetics, extend the range of reconstruction accuracy into the plasma core. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

Theory and application of maximum magnetic energy in toroidal plasmas

International Nuclear Information System (INIS)

Chu, T.K.

1992-02-01

The magnetic energy in an inductively driven steady-state toroidal plasma is a maximum for a given rate of dissipation of energy (Poynting flux). A purely resistive steady state of the piecewise force-free configuration, however, cannot exist, as the periodic removal of the excess poloidal flux and pressure, due to heating, ruptures the static equilibrium of the partitioning rational surfaces intermittently. The rupture necessitates a plasma with a negative q'/q (as in reverse field pinches and spheromaks) to have the same α in all its force-free regions and with a positive q'/q (as in tokamaks) to have centrally peaked α's

Response of plasma rotation to resonant magnetic perturbations in J-TEXT tokamak

Science.gov (United States)

Yan, W.; Chen, Z. Y.; Huang, D. W.; Hu, Q. M.; Shi, Y. J.; Ding, Y. H.; Cheng, Z. F.; Yang, Z. J.; Pan, X. M.; Lee, S. G.; Tong, R. H.; Wei, Y. N.; Dong, Y. B.; J-TEXT Team

2018-03-01

The response of plasma toroidal rotation to the external resonant magnetic perturbations (RMP) has been investigated in Joint Texas Experimental Tokamak (J-TEXT) ohmic heating plasmas. For the J-TEXT’s plasmas without the application of RMP, the core toroidal rotation is in the counter-current direction while the edge rotation is near zero or slightly in the co-current direction. Both static RMP experiments and rotating RMP experiments have been applied to investigate the plasma toroidal rotation. The core toroidal rotation decreases to lower level with static RMP. At the same time, the edge rotation can spin to more than 20 km s-1 in co-current direction. On the other hand, the core plasma rotation can be slowed down or be accelerated with the rotating RMP. When the rotating RMP frequency is higher than mode frequency, the plasma rotation can be accelerated to the rotating RMP frequency. The plasma confinement is improved with high frequency rotating RMP. The plasma rotation is decelerated to the rotating RMP frequency when the rotating RMP frequency is lower than the mode frequency. The plasma confinement also degrades with low frequency rotating RMP.

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

Energy Technology Data Exchange (ETDEWEB)

Kunz, W

1979-01-01

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

Eddy current calculations for the tore supra tokamak

International Nuclear Information System (INIS)

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

1983-01-01

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

A 220 MVA turbo-generator for the TCV tokamak power supplies

International Nuclear Information System (INIS)

Perez, A.; Canay, I.M.; Simond, J.-J.; Morf, J.-J; Pahud, J.-D; Seysen, R.

1989-01-01

A new 220 MVA, 120 Hz, 4 pole turbo-generator will be used as a pulsed power source to supply the toroidal and poloidal power supplies of the TCV tokamak, which is being built at the Ecole Polytechnique Federale de Lausanne in Switzerland. The paper describes the particular requirements of the TCV poloidal power supplies and the main electrical and mechanical features of the turbo-generator and its principal auxillaries. (author). 6 figs.; 1 tab

Relaxation of ion energy spectrum just after turbulent heating pulse in TRIAM-1 tokamak

International Nuclear Information System (INIS)

Nakamura, Kazuo; Hiraki, Naoji; Nakamura, Yukio; Itoh, Satoshi

1982-01-01

The temporal evolution and spatial profile of the ion energy spectrum just after the application of a toroidal current pulse for turbulent heating are investigated experimentally in the TRIAM-1 tokamak and also numerically using the Fokker-Planck equation. The two-component ion energy spectrum formed by turbulent heating relaxes to a single one within tausub(i) (the ion collision time). (author)

Relaxation of ion energy spectrum just after turbulent heating pulse in TRIAM-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

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

1982-07-01

The temporal evolution and spatial profile of the ion energy spectrum just after the application of a toroidal current pulse for turbulent heating are investigated experimentally in the TRIAM-1 tokamak and also numerically using the Fokker-Planck equation. The two-component ion energy spectrum formed by turbulent heating relaxes to a single one within tausub(i) (the ion collision time).

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Modelling of the penetration process of externally applied helical magnetic perturbation of the DED on the TEXTOR tokamak

International Nuclear Information System (INIS)

Kikuchi, Y; Finken, K H; Jakubowski, M; Lehnen, M; Reiser, D; Sewell, G; Wolf, R C

2006-01-01

The error-field penetration process of the dynamic ergodic divertor (DED) on the TEXTOR tokamak has been investigated analytically in terms of a single fluid MHD model with a finite plasma resistivity and viscosity in a cylindrical geometry. The linear model produces a localization of the induced current at the resonance surface and predicts a vortex structure of the velocity field near the resonance layer. Moreover, effects of the Alfven resonance for the error-field penetration are identified by two peaks in the radial profiles of the perturbed toroidal current and the perturbed magnetic flux when the relative rotation velocity between the DED and the rotating tokamak plasma is set to large. Fine structures of the vorticity induced by the DED in the vicinity of the rational surface disappear by introducing a finite plasma perpendicular viscosity. In addition, it is shown that the two peaks of the perturbed toroidal current overlap by an anomalous plasma perpendicular viscosity. Likewise, a bifurcation of the penetration process from the suppressed to the excited state is obtained by a quasi-linear approach taking into account modifications of the radial profiles of the equilibrium current and the plasma rotation due to the DED. A comparison with real experimental results of the DED on the TEXTOR tokamak is shown

Alfven wave heating studies in Tokapole II tokamak

International Nuclear Information System (INIS)

Kortbawi, D.; Witherspoon, F.D.; Zhu, S.Y.; Casavant, T.; Sprott, J.C.; Prager, S.C.

1984-01-01

In earlier experiments at low power on the Tokapole II tokamak using the internal divertor rings as a launching structure the authors have observed a resonance with properties consistent with those expected for a shear Alfven wave. With these encouraging results, a second phase of experiments has begun where, eventually, 4 discrete antennas, located ≅180 0 apart in both the toroidal and poloidal directions and phased to establish proper mode numbers are driven from a 1 MW source. A prototype antenna has been installed and tested. It is a 2 turn Faraday shielded loop extending 54 0 along a toroidal arc. This orientation was chosen for the antenna currents based on the earlier experiments and the simple MHD result that the component of the wage magnetic field perpendicular to the equilibrium field is most strongly divergent. To test this the antenna can be rotated +.45 0 . It can also be inserted radially up to 6 cm

Plasma formation and first OH experiments in GLOBUS-M tokamak

International Nuclear Information System (INIS)

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

2001-01-01

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

3D plasma fluid simulations in divertor tokamaks. Final technical report, 1993--1995

International Nuclear Information System (INIS)

Strauss, H.R.

1995-08-01

The main accomplishment of this grant was the development of a finite element time dependent magnetofluid code, FEMHD. The code is nonlinear and three dimensional. In the poloidal plane, the elemental cells of the mesh are triangles, which offer both simplicity and adaptability. In the third, toroidal, direction, there is an option of a standard staggered finite difference mesh, or Fourier transforms. The FEMHD code runs on several platforms, including Crays, UNIX workstations, and a parallel version runs on an IBM SP1. Several problems have been considered with the unstructured mesh FEMHD code. They are (1) MHD simulations in divertor tokamaks; (2) simulations of ELM-like ballooning modes in divertor tokamaks; and (3) reconnection and singular MHD equilibria

Information content of transient synchrotron radiation in tokamak plasmas

International Nuclear Information System (INIS)

Fisch, N.J.; Kritz, A.H.

1989-04-01

A brief, deliberate, perturbation of hot tokamak electrons produces a transient, synchrotron radiation signal, in frequency-time space, with impressive informative potential on plasma parameters; for example, the dc toroidal electric field, not available by other means, may be measurably. Very fast algorithms have been developed, making tractable a statistical analysis that compares essentially all parameter sets that might possibly explain the transient signal. By simulating data numerically, we can estimate the informative worth of data prior to obtaining it. 20 refs., 2 figs

Gyrokinetic analyses of core heat transport in JT-60U plasmas with different toroidal rotation direction

International Nuclear Information System (INIS)

Narita, Emi; Fukuda, Takeshi; Honda, Mitsuru; Hayashi, Nobuhiko; Urano, Hajime; Ide, Shunsuke

2015-01-01

Tokamak plasmas with an internal transport barrier (ITB) are capable of maintaining improved confinement performance. The ITBs formed in plasmas with the weak magnetic shear and the weak radial electric field shear are often observed to be modest. In these ITB plasmas, it has been found that the electron temperature ITB is steeper when toroidal rotation is in a co-direction with respect to the plasma current than when toroidal rotation is in a counter-direction. To clarify the relationship between the direction of toroidal rotation and heat transport in the ITB region, we examine dominant instabilities using the flux-tube gyrokinetic code GS2. The linear calculations show a difference in the real frequencies; the counter-rotation case has a more trapped electron mode than the co-rotation case. In addition, the nonlinear calculations show that with this difference, the ratio of the electron heat diffusivity χ_e to the ion's χ_i is higher for the counter-rotation case than for the co-rotation case. The difference in χ_e /χ_i agrees with the experiment. We also find that the effect of the difference in the flow shear between the two cases due to the toroidal rotation direction on the linear growth rate is not significant. (author)

Energetic particle drive for toroidicity-induced Alfven eigenmodes and kinetic toroidicity-induced Alfven eigenmodes in a low-shear Tokamak. Revised

International Nuclear Information System (INIS)

Breizman, B.N.; Sharapov, S.E.

1994-10-01

The structure of toroidicity-induced Alfven eigenmodes (TAE) and kinetic TAE (KTAE) with large mode numbers is analyzed and the linear power transfer from energetic particles to these modes is calculated in the low shear limit when each mode is localized near a single gap within an interval whose total width Δ out is much smaller than the radius r m of the mode location. Near its peak where most of the mode energy is concentrated, the mode has an inner scalelength Δ in , which is much smaller than Δ out . The scale Δ in is determined by toroidicity and kinetic effects, which eliminate the singularity of the potential at the resonant surface. This work examines the case when the drift orbit width of energetic particles Δ b is much larger than the inner scalelength Δ in , but arbitrary compared to the total width of the mode. It is shown that the particle-to-wave linear power transfer is comparable for the TAE and KTAE modes in this case. The ratio of the energetic particle contributions to the growth rates of the TAE and KTAE modes is then roughly equal to the inverse ratio of the mode energies. It is found that, in the low shear limit the growth rate of the KTAE modes can be larger than that for the TAE modes

Observation of plasma toroidal-momentum dissipation by neoclassical toroidal viscosity.

Science.gov (United States)

Zhu, W; Sabbagh, S A; Bell, R E; Bialek, J M; Bell, M G; LeBlanc, B P; Kaye, S M; Levinton, F M; Menard, J E; Shaing, K C; Sontag, A C; Yuh, H

2006-06-09

Dissipation of plasma toroidal angular momentum is observed in the National Spherical Torus Experiment due to applied nonaxisymmetric magnetic fields and their plasma-induced increase by resonant field amplification and resistive wall mode destabilization. The measured decrease of the plasma toroidal angular momentum profile is compared to calculations of nonresonant drag torque based on the theory of neoclassical toroidal viscosity. Quantitative agreement between experiment and theory is found when the effect of toroidally trapped particles is included.

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

International Nuclear Information System (INIS)

Kostek, C.; Marshall, T.C.

1982-01-01

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

Use of S-α diagram for representing tokamak equilibrium

International Nuclear Information System (INIS)

Takahashi, H.; Chance, M.; Kessel, C.; LeBlanc, B.; Manickam, J.; Okabayashi, M.

1991-05-01

A use of the S-α diagram is proposed as a tool for representing the plasma equilibrium with a qualitative characterization of its stability through pattern recognition. The diagram is an effective tool for visually presenting the relationship between the shear and dimensionless pressure gradient of an equilibrium. In the PBX-M tokamak, an H-mode operating regime with high poloidal β and L-mode regime with high toroidal β, obtained using different profile modification techniques, are found to have distinct S-α trajectory patterns. Pellet injection into a plasma in the H-mode regime with high toroidal β, obtained using different profile modification techniques, are found to have distinct S-α trajectory patterns. Pellet injection into a plasma in the H-mode regime results in favorable qualities of both regimes. The β collapse process and ELM event also manifest themselves as characteristic changes in the S-α pattern

Realization of toroidal field power supply control system for J-TEXT tokamak

International Nuclear Information System (INIS)

Qiu Shengshun; Zhuang Ge; Zhang Ming; Feng Jianming

2009-01-01

Based on the integrated development environment provided by QNX real-time operation system, the control system of toroidal field power supply is designed and developed. The system is proved to be reliable, stable and in real-time. It can control the power supply successfully to produce a constant current up to 92.5kA lasting for 1s and 1.74T at the magnetic axis. In conclusion, the control system can meet the requirements of the J-TEXT routine operation at present. (authors)

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

International Nuclear Information System (INIS)

Burke, C.

1977-01-01

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

A conceptual design of superconducting spherical tokamak reactor

International Nuclear Information System (INIS)

Nagayama, Yoshio; Shinya, Kichiro; Tanaka, Yasutoshi

2012-01-01

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)

Integrable perturbed magnetic fields in toroidal geometry: An exact analytical flux surface label for large aspect ratio

Energy Technology Data Exchange (ETDEWEB)

Kallinikos, N.; Isliker, H.; Vlahos, L.; Meletlidou, E. [Department of Physics, Aristotle University of Thessaloniki, GR-54124 Thessaloniki (Greece)

2014-06-15

An analytical description of magnetic islands is presented for the typical case of a single perturbation mode introduced to tokamak plasma equilibrium in the large aspect ratio approximation. Following the Hamiltonian structure directly in terms of toroidal coordinates, the well known integrability of this system is exploited, laying out a precise and practical way for determining the island topology features, as required in various applications, through an analytical and exact flux surface label.

Integrable perturbed magnetic fields in toroidal geometry: An exact analytical flux surface label for large aspect ratio

Science.gov (United States)

Kallinikos, N.; Isliker, H.; Vlahos, L.; Meletlidou, E.

2014-06-01

An analytical description of magnetic islands is presented for the typical case of a single perturbation mode introduced to tokamak plasma equilibrium in the large aspect ratio approximation. Following the Hamiltonian structure directly in terms of toroidal coordinates, the well known integrability of this system is exploited, laying out a precise and practical way for determining the island topology features, as required in various applications, through an analytical and exact flux surface label.

Integrable perturbed magnetic fields in toroidal geometry: An exact analytical flux surface label for large aspect ratio

International Nuclear Information System (INIS)

Kallinikos, N.; Isliker, H.; Vlahos, L.; Meletlidou, E.

2014-01-01

An analytical description of magnetic islands is presented for the typical case of a single perturbation mode introduced to tokamak plasma equilibrium in the large aspect ratio approximation. Following the Hamiltonian structure directly in terms of toroidal coordinates, the well known integrability of this system is exploited, laying out a precise and practical way for determining the island topology features, as required in various applications, through an analytical and exact flux surface label

Observation of the low-frequency ion acoustic instability in the turbulently heated TRIAM-1 tokamak plasma

Energy Technology Data Exchange (ETDEWEB)

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

1980-12-01

Density fluctuations in the frequency range of several MHz are observed in the turbulently heated TRIAM-1 tokamak plasma by means of a 4 mm microwave scattering method. It is found from the measurement of the dispersion relation that this instability is considered to be the low-frequency ion acoustic instability propagating nearly perpendicular to the toroidal magnetic field.

Parallel gradient effects on ICRH (Ion Cyclotron Resonance Heating) in Tokamaks

International Nuclear Information System (INIS)

Smithe, D.N.

1987-01-01

This dissertation examines the effects on Ion Cyclotron Resonance Heating of parallel nonuniformity in the magnetic field which arises from the poloidal field in a tokamak and the universal (major radius)/sup /minus/1/ scaling of the cyclotron frequency. The goal of the analysis is the macroscopic warm plasma current including temperature in the sense of the finite Larmor radius expansion and the quasilocal approximation of the parallel guiding center motion. A 1-D numerical application of the fully nonlocal integral dielectric is performed. Parallel gradient effects are studied for He-3 minority, 2nd harmonic deuterium, and hydrogen minority heating in tokamaks. The results show quite significant alteration of the toroidal wavenumber absorption spectrum, and a wealth of new behavior on the local propagation scale. 95 refs., 37 figs

Physics issues of high bootstrap current tokamaks

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

Bornatici, M.; Engelmann, F.

1978-01-01

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

Tokamak βaB/I limit and its dependence on the safety factor

International Nuclear Information System (INIS)

Ramos, J.J.

1990-01-01

It is shown that, for a given tokamak cross-sectional shape and arbitrary values of the magnetic axis safety factor q 0 , the first stability condition against pressure-driven magnetohydrodynamic modes has the form 40πβaB 0 /I≤C R (q 0 )/q 0 . Moreover, in the limit of large q 0 , C R (q 0 ) becomes independent of q 0 and independent of the toroidal mode number

Neoclassical transport of impurtities in tokamak plasmas

International Nuclear Information System (INIS)

Hirshman, S.P.; Sigmar, D.J.

1981-05-01

Tokamak plasmas are inherently comprised of multiple ion species. This is due to wall-bred impurities and, in future reactors, will result from fusion-born alpha particles. Relatively small concentrations of highly charged non-hydrogenic impurities can strongly influence plasma transport properties whenever n/sub I/e/sub I/ 2 /n/sub H/e 2 greater than or equal to (m/sub e//m/sub H/)/sup 1/2/. The determination of the complete neoclassical Onsager matrix for a toroidally confined multispecies plasma, which provides the linear relation between the surface averaged radial fluxes and the thermodynamic forces (i.e., gradients of density and temperature, and the parallel electric field), is reviewed. A closed set of one-dimensional moment equations is presented for the time evolution of thermodynamic and magnetic field quantities which results from collisional transport of the plasma and two dimensional motion of the magnetic flux surface geometry. The effects of neutral beam injection on the equilibrium and transport properties of a toroidal plasma are consistently included

Plasma residual poloidal rotation in TCABR tokamak

International Nuclear Information System (INIS)

Severo, J.H.F.; Nascimento, I.C.; Tsypin, V.S.; Galvao, R.M.O.

2003-01-01

This paper reports the first measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak for a collisional plasma (Pfirsch-Schluter regime), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5±1.0)·10 3 m/s at r ∼ 2/3a, (a - limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) · 10 3 m/s for the plasma core to a co-current value of (2.0 ± 1.0) · 10 3 m/s near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner. (author)

Midplane Faraday rotation: A densitometer for large tokamaks