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

Topology of tokamak orbits

International Nuclear Information System (INIS)

Rome, J.A.; Peng, Y.K.M.

1978-09-01

Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

Study of heat fluxes on plasma facing components in a tokamak from measurements of temperature by infrared thermography

International Nuclear Information System (INIS)

Daviot, R.

2010-05-01

The goal of this thesis is the development of a method of computation of those heat loads from measurements of temperature by infrared thermography. The research was conducted on three issues arising in current tokamaks but also future ones like ITER: the measurement of temperature on reflecting walls, the determination of thermal properties for deposits observed on the surface of tokamak components and the development of a three-dimensional, non-linear computation of heat loads. A comparison of several means of pyrometry, monochromatic, bi-chromatic and photothermal, is performed on an experiment of temperature measurement. We show that this measurement is sensitive to temperature gradients on the observed area. Layers resulting from carbon deposition by the plasma on the surface of components are modeled through a field of equivalent thermal resistance, without thermal inertia. The field of this resistance is determined, for each measurement points, from a comparison of surface temperature from infrared thermographs with the result of a simulation, which is based on a mono-dimensional linear model of components. The spatial distribution of the deposit on the component surface is obtained. Finally, a three-dimensional and non-linear computation of fields of heat fluxes, based on a finite element method, is developed here. Exact geometries of the component are used. The sensitivity of the computed heat fluxes is discussed regarding the accuracy of the temperature measurements. This computation is applied to two-dimensional temperature measurements of the JET tokamak. Several components of this tokamak are modeled, such as tiles of the divertor, upper limiter and inner and outer poloidal limiters. The distribution of heat fluxes on the surface of these components is computed and studied along the two main tokamak directions, poloidal and toroidal. Toroidal symmetry of the heat loads from one tile to another is shown. The influence of measurements spatial resolution

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.)

Magnetic flux reconstruction methods for shaped tokamaks

International Nuclear Information System (INIS)

Tsui, Chi-Wa.

1993-12-01

The use of a variational method permits the Grad-Shafranov (GS) equation to be solved by reducing the problem of solving the 2D non-linear partial differential equation to the problem of minimizing a function of several variables. This high speed algorithm approximately solves the GS equation given a parameterization of the plasma boundary and the current profile (p' and FF' functions). The author treats the current profile parameters as unknowns. The goal is to reconstruct the internal magnetic flux surfaces of a tokamak plasma and the toroidal current density profile from the external magnetic measurements. This is a classic problem of inverse equilibrium determination. The current profile parameters can be evaluated by several different matching procedures. Matching of magnetic flux and field at the probe locations using the Biot-Savart law and magnetic Green's function provides a robust method of magnetic reconstruction. The matching of poloidal magnetic field on the plasma surface provides a unique method of identifying the plasma current profile. However, the power of this method is greatly compromised by the experimental errors of the magnetic signals. The Casing Principle provides a very fast way to evaluate the plasma contribution to the magnetic signals. It has the potential of being a fast matching method. The performance of this method is hindered by the accuracy of the poloidal magnetic field computed from the equilibrium solver. A flux reconstruction package has been implemented which integrates a vacuum field solver using a filament model for the plasma, a multi-layer perception neural network as an interface, and the volume integration of plasma current density using Green's functions as a matching method for the current profile parameters. The flux reconstruction package is applied to compare with the ASEQ and EFIT data. The results are promising

Feasibility of laser pumping with neutron fluxes from present-day large tokamaks

Energy Technology Data Exchange (ETDEWEB)

Jassby, D.L.

1986-08-01

The minimum fusion-neutron flux needed to observe nuclear-pumped lasing with tokamaks can be reduced substantially by optimizing neutron scattering into the laser cell, located between adjacent toroidal-field coils. The laser lines most readily pumped are probably the /sup 3/He-Ne lines at 0.633 ..mu.. and in the infrared, where the /sup 3/He-Ne gas is excited by energetic ions produced in the /sup 3/He(n,p)T reaction. These lines are expected to lase at the levels of D-T neutron flux foreseen for the TFTR in 1989 (>>10/sup 12/ n/cm/sup 2//s), while amplification should be observable at the existing levels of D-D neutron flux (greater than or equal to 5 x 10/sup 9/ n/cm/sup 2//s). Lasing on the 1.73 ..mu.. and 2.63 ..mu.. transitions of Xe may be observable at the maximum expected levels of D-T neutron flux in TFTR enhanced by scattering.

Feasibility of laser pumping with neutron fluxes from present-day large tokamaks

International Nuclear Information System (INIS)

Jassby, D.L.

1986-08-01

The minimum fusion-neutron flux needed to observe nuclear-pumped lasing with tokamaks can be reduced substantially by optimizing neutron scattering into the laser cell, located between adjacent toroidal-field coils. The laser lines most readily pumped are probably the 3 He-Ne lines at 0.633 μ and in the infrared, where the 3 He-Ne gas is excited by energetic ions produced in the 3 He(n,p)T reaction. These lines are expected to lase at the levels of D-T neutron flux foreseen for the TFTR in 1989 (>>10 12 n/cm 2 /s), while amplification should be observable at the existing levels of D-D neutron flux (≥ 5 x 10 9 n/cm 2 /s). Lasing on the 1.73 μ and 2.63 μ transitions of Xe may be observable at the maximum expected levels of D-T neutron flux in TFTR enhanced by scattering

Impurity flux collection at the plasma edge of the tokamak MT-1

International Nuclear Information System (INIS)

Hildebrandt, D.; Bakos, J.S.; Petravich, G.

1989-09-01

Fluxes of intrinsic and injected impurities and background plasma ions were collected using a bidirectional probe at the plasma edge of the tokamak MT-1. The directional and radial dependences of injected impurities and plasma ions were very similar indicating a strong coupling of the impurity transport to the dynamics of the background plasma. The measured intrinsic concentration of about 10 -4 for Mo at the plasma edge is derived. (author) 17 refs.; 5 figs

Volt-second consumption in tokamaks with sawtooth activity

International Nuclear Information System (INIS)

Houlberg, W.A.

1987-01-01

The effects of sawtooth activity on the poloidal magnetic flux and energy balances in tokamak plasmas on a diffusive timescale are evaluated through the application of conservation principles to Maxwell's equations. Poloidal magnetic flux (volt-second) consumption can be partitioned into internal and dissipative components by two methods: the 'axial method' based on a magnetic flux balance and the 'Poynting method' based on a magnetic energy balance. Both require additional terms that specifically account for the poloidal flux and magnetic energy changes during magnetic reconnection derived from analysis on a magnetohydrodynamic (MHD) timescale. In experimental analyses these terms are absorbed in the inferred resistive dissipation, while in predictive analyses thay must be evaluated directly. The dissipation of poloidal flux by sawtooth activity can exceed the normal resistive dissipation when the axial method of accounting is used

Disruption-induced poloidal currents in the tokamak wall

International Nuclear Information System (INIS)

Pustovitov, V.D.

2017-01-01

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

Disruption-induced poloidal currents in the tokamak wall

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-15

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

Experience with high heat flux components in large tokamaks

International Nuclear Information System (INIS)

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

1991-01-01

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

Stability of force-free spheromak plasma in spheroidal flux conserver

International Nuclear Information System (INIS)

Kaneko, Shobu; Tsutsui, Hiroaki

1988-01-01

The Woltjer-Taylor method is applied to spheromak plasmas in spheroidal flux conservers. As models of the flux conserver, both oblate and prolate spheroidal vessels with a center conductor are used. The plasma is not assumed to be nearly spherical, and the Rayleigh-Ritz method and the finite element method are used to evaluate the eigenvalues. The oblate spheromak is shown to be stable irrespective of the shape of the flux conserver. Though the prolate spheromak is unstable if there is no center conductor, it can be stable if the center conductor is installed. (author)

Effects of flux conservation on the field configuration in Scyllac

International Nuclear Information System (INIS)

Van der Laan, P.C.T.

1977-04-01

Flux conservation in Scyllac-type experiments shows up in two ways. First of all the poloidal flux between the outside edge of the plasma and the inside of the coil is conserved. This requires a net longitudinal current in the plasma, to cancel the poloidal flux caused by the helical stellarator fields. An expression for this net current is derived, and effects that could occur in sector experiments are discussed. The flux conservation inside the conducting plasma leads to a conservation of the local rotational transform. Since the pinch itself is surrounded by a well-conducting low-density plasma, the rotational transform is conserved in a wide region. Depending on the time history of the applied fields, volume currents are induced in this region, as is shown for two examples. Although an additional capacitor bank can be used to cancel the net current, a cancellation of all the volume currents is extremely difficult. The resulting equilibrium configurations differ considerably from the Scyllac equilibria without volume currents, which are used in stability calculations

Magnetohydrodynamic stability of spheromak plasma in spheroidal flux conserver

International Nuclear Information System (INIS)

Kaneko, Shobu; Kamitani, Atsushi.

1985-11-01

The MHD equilibrium configurations of spheromak plasmas in a spheroidal flux conserver are determined by use of a pressure distribution whose derivative dp/dψ vanishes on the magnetic axis, and by use of an optimized distribution. Here p is the pressure and ψ is the flux function. These equilibria are shown to be stable for symmetric modes. The stability for localized modes is investigated by the Mercier criterion. The values of the maximum beta ratio β max are evaluated for both pressure distributions and are shown to become about two times larger by optimization. If the condition, q axis max are found to be less than 30 %. The oblate spheroidal flux conserver is shown to be better than the toroidal conserver with a rectangular cross section from the standpoint of stability. (author)

Computational study of axisymmetric modes in noncircular cross section tokamaks

International Nuclear Information System (INIS)

Johnson, J.L.; Chance, M.S.; Greene, J.M.; Grimm, R.C.; Jardin, S.C.; Kerner, W.; Manickam, J.; Weimer, K.E.

1976-09-01

A major computational program to investigate the MHD equilibrium, stability, and nonlinear evolution properties of realistic tokamak configurations is proceeding. Preliminary application is made to the Princeton PDX device. Both axisymmetric (n = 0) modes and kink (n = 1) modes are found; the growth rates depend sensitively on the configuration. A study of the nonlinear evolution of axisymmetric modes in such a device shows that flux conservation in the vacuum region can limit their growth

State diagrams of tokamaks and state transitions

International Nuclear Information System (INIS)

Minardi, E.

1992-01-01

In a simple one-fluid cylindrical model of transport and of dissipative effects, the family of the magnetic states of the Tokamak which correspond to a vanishing entropy production in the confinement region is characterized by a define relation or ''state equation'' involving the relevant parameters of the discharge. An investigation is made as to how the entropy production changes when the current density profile is rearranged by a perturbation which conserves the poloidal magnetic flux. It is shown that for a sufficiently short time interval, that is to say t 2 E τ s where τ E is the energy confinement time and τ s is the resistive time, neighbouring bifurcating equilibria exist which can be reached with a flux-conserving transition and with increase of the magnetic entropy. The family of these new states can also be characterized by a state equation involving the relevant discharge parameters. When the state equations of the two families are simultaneously satisfied by the same set of parameter values, a flux-conserving, entropy-increasing transition may take place between states of the two families. The modifications of the current density and of the temperature profiles involved in the transition and the conditions that the discharge parameters should satisfy in order that the transition could occur are investigated. (author)

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.

Preliminary oscillating fluxes current drive experiment in DIII-D tokamak

International Nuclear Information System (INIS)

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

1995-01-01

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

The tokamak as a neutron source

International Nuclear Information System (INIS)

Hendel, H.W.; Jassby, D.L.

1989-11-01

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

Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

International Nuclear Information System (INIS)

Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

2015-01-01

Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor

Plasma column displacement measurements by modified Rogowski sine-coil and Biot-Savart/magnetic flux equation solution on IR-T1 tokamak

International Nuclear Information System (INIS)

Razavi, M.; Mollai, M.; Khorshid, P.; Nedzelskiy, I.; Ghoranneviss, M.

2010-01-01

The modified Rogowski sine-coil (MRSC) has been designed and implemented for the plasma column horizontal displacement measurements on small IR-T1 tokamak. MRSC operation has been examined on test assembly and tokamak. Obtained results show high sensitivity to the plasma column horizontal displacement and negligible sensitivity to the vertical displacement; linearity in wide, ±0.1 m, range of the displacements; and excellent, 1.5%, agreement with the results of numerical solution of Biot-Savart and magnetic flux equations.

Investigation of edge neutral flux on the ISX-B tokamak using a low-energy charge-exchange analyzer

Energy Technology Data Exchange (ETDEWEB)

Thomas, D. M.

1983-08-01

To study the emission of D/sup 0/ from the periphery of a tokamak plasma, a low-energy neutral particle spectrometer optimized for (16 < E < 500 eV) has been built and employed on the Impurity Study Experiment (ISX-B) tokamak. The diagnostic utilizes a cesium vapor cell to form negative ions from the incident D/sup 0/ neutrals and a four-channel electrostatic analyzer to energy analyze the negative ions. The spectrometer was absolutely calibrated using D/sup 0/ beams formed by electron capture by positive ions in a gas cell and by photo-detachment of negative ions by a yttrium-aluminum-garnet laser. For the observation region chosen on ISX-B (120/sup 0/ toroidally away from the limiter, near the gas puff), the neutral particle flux has a two-component nature. These data are well fit by two separate exponential distributions of equivalent temperatures 6 to 8 eV for particle energies below about 80 eV and 70 to 80 eV for particle energies above 80 eV. For ohmically heated discharges, the measured particle flux in the energy range 25 to 700 eV is approx. 2.5 x 10/sup 15/ cm/sup -2/.s/sup -1/; the mean particle energy is approx. 70 eV, and the calculated flux at the wall is approx. 30 mW/cm/sup 2/. The major effect of neutral beam heating is to increase the particle flux in the 25- to 700-eV range by a factor of 3.

Motion of a compact toroid inside a cylindrical flux conserver

Energy Technology Data Exchange (ETDEWEB)

Jarboe, T.R.; Henins, I.; Hoida, H.W.; Linford, R.K.; Marshall, J.; Platts, D.A.; Sherwood, A.R.

1980-10-13

Compact toroids have been generated in a cylindrical resistive flux conserver. They are observed to rotate so that their major axis is perpendicular to the axis of the flux conserver. Subsequently they remain stationary and their magnetic fields decay with a time constant of about 100 ..mu..s. This is the first observation of the predicted tipping mode and its saturation when no external fields are present. The compact toroids contain toroidal fields and are initially prolate in shape.

Motion of a compact toroid inside a cylindrical flux conserver

International Nuclear Information System (INIS)

Jarboe, T.R.; Henins, I.; Hoida, H.W.; Linford, R.K.; Marshall, J.; Platts, D.A.; Sherwood, A.R.

1980-01-01

Compact toroids have been generated in a cylindrical resistive flux conserver. They are observed to rotate so that their major axis is perpendicular to the axis of the flux conserver. Subsequently they remain stationary and their magnetic fields decay with a time constant of about 100 μs. This is the first observation of the predicted tipping mode and its saturation when no external fields are present. The compact toroids contain toroidal fields and are initially prolate in shape

Stochastic line motion and stochastic flux conservation for nonideal hydromagnetic models

International Nuclear Information System (INIS)

Eyink, Gregory L.

2009-01-01

We prove that smooth solutions of nonideal (viscous and resistive) incompressible magnetohydrodynamic (MHD) equations satisfy a stochastic law of flux conservation. This property implies that the magnetic flux through a surface is equal to the average of the magnetic fluxes through an ensemble of surfaces advected backward in time by the plasma velocity perturbed with a random white noise. Our result is an analog of the well-known Alfven theorem of ideal MHD and is valid for any value of the magnetic Prandtl number. A second stochastic conservation law is shown to hold at unit Prandtl number, a random version of the generalized Kelvin theorem derived by Bekenstein and Oron for ideal MHD. These stochastic conservation laws are not only shown to be consequences of the nonideal MHD equations but are proved in fact to be equivalent to those equations. We derive similar results for two more refined hydromagnetic models, Hall MHD and the two-fluid plasma model, still assuming incompressible velocities and isotropic transport coefficients. Finally, we use these results to discuss briefly the infinite-Reynolds-number limit of hydromagnetic turbulence and to support the conjecture that flux conservation remains stochastic in that limit.

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

International Nuclear Information System (INIS)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

2004-01-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

Energy Technology Data Exchange (ETDEWEB)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

2004-09-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Modelling dust transport in tokamaks

International Nuclear Information System (INIS)

Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

2008-01-01

The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

Feasibility study of fast swept divertor strike point suppressing transient heat fluxes in big tokamaks.

Czech Academy of Sciences Publication Activity Database

Horáček, Jan; Cunningham, G.; Entler, Slavomír; Dobias, P.; Duban, R.; Imríšek, Martin; Markovič, Tomáš; Havlíček, Josef; Enikeev, R.

2017-01-01

Roč. 123, November (2017), s. 646-649 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) LM2015045; GA MŠk(CZ) 8D15001; GA MŠk LG14002 Institutional support: RVO:61389021 Keywords : DEMO * ELM * Divertor * Heat flux * Tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617300376

Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

International Nuclear Information System (INIS)

Giannone, L.; Schneider, W.; McCarthy, P.J.; Sips, A.C.C.; Treutterer, W.; Behler, K.; Eich, T.; Fuchs, J.C.; Hicks, N.; Kallenbach, A.; Maraschek, M.; Mlynek, A.; Neu, G.; Pautasso, G.; Raupp, G.; Reich, M.; Schuhbeck, K.H.; Stober, J.; Volpe, F.; Zehetbauer, T.

2009-01-01

The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

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

International Nuclear Information System (INIS)

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

1980-03-01

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

The influence of plasma horizontal position on the neutron rate and flux of neutral atoms in injection heating experiment on the TUMAN-3M tokamak

Science.gov (United States)

Kornev, V. A.; Chernyshev, F. V.; Melnik, A. D.; Askinazi, L. G.; Wagner, F.; Vildjunas, M. I.; Zhubr, N. A.; Krikunov, S. V.; Lebedev, S. V.; Razumenko, D. V.; Tukachinsky, A. S.

2013-11-01

Horizontal displacement of plasma along the major radius has been found to significantly influence the fluxes of 2.45 MeV DD neutrons and high-energy charge-exchange atoms from neutral beam injection (NBI) heated plasma of the TUMAN-3M tokamak. An inward shift by Δ R = 1 cm causes 1.2-fold increase in the neutron flux and 1.9-fold increase in the charge-exchange atom flux. The observed increase in the neutron flux is attributed to joint action of several factors-in particular, improved high-energy ion capture and confinement and, probably, decreased impurity inflow from the walls, which leads to an increase in the density of target ions. A considerable increase in the flux of charge-exchange neutrals in inward-shifted plasma is due to the increased number of captured high-energy ions and, to some extent, the increased density of the neutral target. As a result of the increase in the content of high-energy ions, the central ion temperature T i (0) increased from 250 to 350 eV. The dependence of the neutron rate on major radius R 0 should be taken into account when designing compact tokamak-based neutron sources.

Multiscale gyrokinetics for rotating tokamak plasmas: fluctuations, transport and energy flows.

Science.gov (United States)

Abel, I G; Plunk, G G; Wang, E; Barnes, M; Cowley, S C; Dorland, W; Schekochihin, A A

2013-11-01

This paper presents a complete theoretical framework for studying turbulence and transport in rapidly rotating tokamak plasmas. The fundamental scale separations present in plasma turbulence are codified as an asymptotic expansion in the ratio ε = ρi/α of the gyroradius to the equilibrium scale length. Proceeding order by order in this expansion, a set of coupled multiscale equations is developed. They describe an instantaneous equilibrium, the fluctuations driven by gradients in the equilibrium quantities, and the transport-timescale evolution of mean profiles of these quantities driven by the interplay between the equilibrium and the fluctuations. The equilibrium distribution functions are local Maxwellians with each flux surface rotating toroidally as a rigid body. The magnetic equilibrium is obtained from the generalized Grad-Shafranov equation for a rotating plasma, determining the magnetic flux function from the mean pressure and velocity profiles of the plasma. The slow (resistive-timescale) evolution of the magnetic field is given by an evolution equation for the safety factor q. Large-scale deviations of the distribution function from a Maxwellian are given by neoclassical theory. The fluctuations are determined by the 'high-flow' gyrokinetic equation, from which we derive the governing principle for gyrokinetic turbulence in tokamaks: the conservation and local (in space) cascade of the free energy of the fluctuations (i.e. there is no turbulence spreading). Transport equations for the evolution of the mean density, temperature and flow velocity profiles are derived. These transport equations show how the neoclassical and fluctuating corrections to the equilibrium Maxwellian act back upon the mean profiles through fluxes and heating. The energy and entropy conservation laws for the mean profiles are derived from the transport equations. Total energy, thermal, kinetic and magnetic, is conserved and there is no net turbulent heating. Entropy is produced

Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

Energy Technology Data Exchange (ETDEWEB)

Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

2015-09-15

The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

Fast IR diodes thermometer for tokamak

International Nuclear Information System (INIS)

Chen Xiangbo

2001-01-01

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

Heat load material studies: Simulated tokamak disruptions

International Nuclear Information System (INIS)

Gahl, J.M.; McDonald, J.M.; Zakharov, A.; Tserevitinov, S.; Barabash, V.; Guseva, M.

1991-01-01

It is clear that an improved understanding of the effects of tokamak disruptions on plasma facing component materials is needed for the ITER program. very large energy fluxes are predicted to be deposited in ITER and could be very damaging to the machine. During 1991, Sandia National Laboratories and the University of New Mexico conducted cooperative tokamak disruption simulation experiments at several Soviet facilities. These facilities were located at the Efremov Institute in Leningrad, the Kurchatov Atomic Energy Institute (Troisk and Moscow) and the Institute for Physical Chemistry of the Soviet Adademy of Sciences in Moscow. Erosion of graphite from plasma stream impact is seen to be much less than that observed with laser or electron beams with similar energy fluxes. This, along with other data obtained, seem to suggest that the ''vapor shielding'' effect is a very important phenomenon in the study of graphite erosion during tokamak disruption

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

On steady poloidal and toroidal flows in tokamak plasmas

International Nuclear Information System (INIS)

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

2010-01-01

The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B θ /B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B θ /B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.

The energy principle applied to diverted tokamak configurations

International Nuclear Information System (INIS)

Atanasiu, C. V.; Guenter, S.; Lackner, K.; Moraru, A.; Zakharov, L. E.; Subbotin, A. A.

2008-01-01

Writing the expression of the potential energy in terms of the perturbation of the flux function, and performing an Euler minimisation, one obtains a system of ordinary differential equations in that perturbation. For a diverted configuration, the usual vanishing boundary conditions for the perturbed flux function at the magnetic axis and at infinity can no longer be used. In place of the vanishing boundary conditions at infinity, an approach to fix 'natural' boundary conditions for the system of differential equations for the perturbed flux function, just at the plasma boundary has been developed. As an example of application of our approaches, a particular equilibrium configuration of the ASDEX Upgrade tokamak has been considered and a detailed investigation of the dependence of the tearing stability parameter Δ' on plasma shape is given for a realistic tokamak equilibrium. The results shown are at least in qualitative agreement with experimental observations on ASDEX Upgrade and JET of a stabilizing influence of triangularity. The knowledge of Δ' for realistic tokamak plasmas is especially important for understanding of the plasma stability against NTMs. (authors)

Neoclassical transport analysis for a class of high-β tokamak equilibria

International Nuclear Information System (INIS)

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

1995-01-01

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

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.)

Continuum analysis of biological systems conserved quantities, fluxes and forces

CERN Document Server

Suraishkumar, G K

2014-01-01

This book addresses the analysis, in the continuum regime, of biological systems at various scales, from the cellular level to the industrial one. It presents both fundamental conservation principles (mass, charge, momentum and energy) and relevant fluxes resulting from appropriate driving forces, which are important for the analysis, design and operation of biological systems. It includes the concept of charge conservation, an important principle for biological systems that is not explicitly covered in any other book of this kind. The book is organized in five parts: mass conservation; charge conservation; momentum conservation; energy conservation; and multiple conservations simultaneously applied. All mathematical aspects are presented step by step, allowing any reader with a basic mathematical background (calculus, differential equations, linear algebra, etc.) to follow the text with ease. The book promotes an intuitive understanding of all the relevant principles and in so doing facilitates their applica...

Gyrokinetic theory and dynamics of the tokamak edge

Energy Technology Data Exchange (ETDEWEB)

Scott, B. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

2016-08-15

The validity of modern gyrokinetic field theory is assessed for the tokamak edge. The basic structure of the Lagrangian and resulting equations and their conservation laws is reviewed. The conventional microturbulence ordering for expansion is small potential/arbitrary wavelength. The equilibrium ordering for expansion is long wavelength/arbitrary amplitude. The long-wavelength form of the conventional Lagrangian is derived in detail. The two Lagrangians are shown to match at long wavelength if the E x B Mach number is small enough for its corrections to the gyroaveraging to be neglected. Therefore, the conventional derivation and its Lagrangian can be used at all wavelengths if these conditions are satisfied. Additionally, dynamical compressibility of the magnetic field can be neglected if the plasma beta is small. This allows general use of a shear-Alfven Lagrangian for edge turbulence and self consistent equilibrium-scale phenomena for flows, currents, and heat fluxes for conventional tokamaks without further modification by higher-order terms. Corrections in polarisation and toroidal angular momentum transport due to these higher-order terms for global edge turbulence computations are shown to be small. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Ballooning mode second stability region for sequences of tokamak equilibria

International Nuclear Information System (INIS)

Sugiyama, L.; Mark, J.W.K.

A numerical study of several sequences of tokamak equilibria derived from two flux conserving sequences confirms the tendency of high n ideal MHD ballooning modes to stabilize for values of the plasma beta greater than a second critical beta, for sufficiently favorable equilibria. The major stabilizing effect of increasing the inverse rotational transform profile q(Psi) for equilibria with the same flux surface geometry is shown. The unstable region shifts toward larger shear d ln q/d ln γ and the width of the region measured in terms of the poloidal beta or a pressure gradient parameter, for fixed shear, decreases. The smaller aspect ratio sequences are more sensitive to changes in q and have less stringent limits on the attainable value of the plasma beta in the high beta stable region. Finally, the disconnected mode approximation is shown to provide a reasonable description of the second high beta stability boundary

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

Fusion-product transport in axisymmetric tokamaks: losses and thermalization

International Nuclear Information System (INIS)

Hively, L.M.

1980-01-01

High-energy fusion-product losses from an axisymmetric tokamak plasma are studied. Prompt-escape loss fluxes (i.e. prior to slowing down) are calculated including the non-separable dependence of flux as a function of poloidal angle and local angle-of-incidence at the first wall. Fusion-product (fp) thermalization and heating are calculated assuming classical slowing down. The present analytical model describes fast ion orbits and their distribution function in realistic, high-β, non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated

Control of three dimensional particle flux to divertor using rotating RMP in the EAST tokamak

Science.gov (United States)

Jia, M.; Sun, Y.; Liang, Y.; Wang, L.; Xu, J.; Gu, S.; Lyu, B.; Wang, H. H.; Yang, X.; Zhong, F.; Chu, N.; Feng, W.; He, K.; Liu, Y. Q.; Qian, J.; Shi, T.; Shen, B.

2018-04-01

Controlling the steady state particle and heat flux impinging on the plasma facing components, as one of the main concerns of future fusion reactors, is still necessary when the transient power loads induced by edge localized modes (ELMs) have been eliminated by resonant magnetic perturbations (RMPs) in high confinement tokamak experiments. This is especially true for long pulse operation. One promising solution is to use the rotating perturbed field. Recently rotating and differential phase scans of n  =  1 and 2 RMP fields have been operated for the first time in EAST discharges. The particle flux patterns on the divertor targets change synchronously with both rotating and phasing RMP fields as predicted by the modeled magnetic footprint patterns. The modeling with plasma response, which is calculated by MARS-F, is also carried out. The plasma response shows amplifying or screening effect to n  =  2 perturbations with different spectra. This changes the field line penetration depth rather than the general footprint shape. This has been verified by experimental observations on EAST. These experiments motivate further study of reducing both transient and steady state local power load and particle flux with the help of rotating RMPs in long pulse operation.

Visible imaging measurement of position and displacement of the last closed flux surface in EAST tokamak

Energy Technology Data Exchange (ETDEWEB)

Wang, Y.F. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Xu, G.S., E-mail: gsxu@ipp.ac.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Li, Y.L.; Yang, J.H.; Yan, N.; Liu, L.; Yuan, S.; Luo, Z.P. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sang, C.F. [School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Gu, S.; Xu, J.C.; Hu, G.H.; Wang, Y.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Peng, Y.K.M.; Wan, B.N. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2017-06-15

Highlights: • A new method for measuring the position and displacement of the LCFS has been developed in EAST tokamak. • This method is based on the visible imaging diagnostic and shown to be an effective and convenient approach. • This method can be applied to measure displacements of the LCFS during application of resonant magnetic perturbation fields. - Abstract: A new method for measuring the position and displacement of the last closed flux surface (LCFS) with visible imaging diagnostics has been developed in EAST. By measuring the relative intensity profiles of the green visible Li-II emission in the tangential planes of the optical systems, it is possible to infer the positions of certain points on the LCFS. This emission line is readily available in discharges with Li-coating wall routinely employed to improve the plasma performance. We describe the measuring method, giving results which are compared with those obtained by EFIT, and showing this as an effective and convenient approach to determine the position of the LCFS. This method is further applied to measure the displacements of the LCFS during application of resonant magnetic perturbation fields in the EAST tokamak.

Flux continuity and probability conservation in complexified Bohmian mechanics

International Nuclear Information System (INIS)

Poirier, Bill

2008-01-01

Recent years have seen increased interest in complexified Bohmian mechanical trajectory calculations for quantum systems as both a pedagogical and computational tool. In the latter context, it is essential that trajectories satisfy probability conservation to ensure they are always guided to where they are most needed. We consider probability conservation for complexified Bohmian trajectories. The analysis relies on time-reversal symmetry considerations, leading to a generalized expression for the conjugation of wave functions of complexified variables. This in turn enables meaningful discussion of complexified flux continuity, which turns out not to be satisfied in general, though a related property is found to be true. The main conclusion, though, is that even under a weak interpretation, probability is not conserved along complex Bohmian trajectories

Including collisions in gyrokinetic tokamak and stellarator simulations

International Nuclear Information System (INIS)

Kauffmann, Karla

2012-01-01

Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the δf method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal ITGs and

Including collisions in gyrokinetic tokamak and stellarator simulations

Energy Technology Data Exchange (ETDEWEB)

Kauffmann, Karla

2012-04-10

Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal

Boundary plasma heat flux width measurements for poloidal magnetic fields above 1 Tesla in the Alcator C-Mod tokamak

Science.gov (United States)

Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Alcator C-Mod Team

2017-10-01

The boundary heat flux width, along with the total power flowing into the boundary, sets the power exhaust challenge for tokamaks. A multi-machine boundary heat flux width database found that the heat flux width in H-modes scaled inversely with poloidal magnetic field (Bp) and was independent of machine size. The maximum Bp in the database was 0.8 T, whereas the ITER 15 MA, Q =10 scenario will be 1.2 T. New measurements of the boundary heat flux width in Alcator C-Mod extend the international database to plasmas with Bp up to 1.3 T. C-Mod was the only experiment able to operate at ITER-level Bp. These new measurements are from over 300 plasma shots in L-, I-, and EDA H-modes spanning essentially the whole operating space in C-Mod. We find that the inverse-Bp dependence of the heat flux width in H-modes continues to ITER-level Bp, further reinforcing the empirical projection of 500 μm heat flux width for ITER. We find 50% scatter around the inverse-Bp scaling and are searching for the `hidden variables' causing this scatter. Supported by USDoE award DE-FC02-99ER54512.

Investigation of edge neutral flux on the ISX-B tokamak using a low-energy charge-exchange analyzer

International Nuclear Information System (INIS)

Thomas, D.M.

1983-08-01

To study the emission of D 0 from the periphery of a tokamak plasma, a low-energy neutral particle spectrometer optimized for (16 0 neutrals and a four-channel electrostatic analyzer to energy analyze the negative ions. The spectrometer was absolutely calibrated using D 0 beams formed by electron capture by positive ions in a gas cell and by photo-detachment of negative ions by a yttrium-aluminum-garnet laser. For the observation region chosen on ISX-B (120 0 toroidally away from the limiter, near the gas puff), the neutral particle flux has a two-component nature. These data are well fit by two separate exponential distributions of equivalent temperatures 6 to 8 eV for particle energies below about 80 eV and 70 to 80 eV for particle energies above 80 eV. For ohmically heated discharges, the measured particle flux in the energy range 25 to 700 eV is approx. 2.5 x 10 15 cm - 2 .s - 1 ; the mean particle energy is approx. 70 eV, and the calculated flux at the wall is approx. 30 mW/cm 2 . The major effect of neutral beam heating is to increase the particle flux in the 25- to 700-eV range by a factor of 3

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

The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

Energy Technology Data Exchange (ETDEWEB)

Harrison, J. R.; Fishpool, G. M.; Thornton, A. J. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Walkden, N. R. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

2015-09-15

The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFR of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.

Development of Multi-Dimensional RELAP5 with Conservative Momentum Flux

Energy Technology Data Exchange (ETDEWEB)

Jang, Hyung Wook; Lee, Sang Yong [KINGS, Ulsan (Korea, Republic of)

2016-10-15

The non-conservative form of the momentum equations are used in many codes. It tells us that using the non-conservative form in the non-porous or open body problem may not be good. In this paper, two aspects concerning the multi-dimensional codes will be discussed. Once the validity of the modified code is confirmed, it is applied to the analysis of the large break LOCA for APR-1400. One of them is the properness of the type of the momentum equations. The other discussion will be the implementation of the conservative momentum flux term in RELAP5. From the present study and former, it is shown that the RELAP5 Multi-D with conservative convective terms is applicable to LOCA analysis. And the implementation of the conservative convective terms in RELAP5 seems to be successful. Further efforts have to be made on making it more robust.

Evolution of the transport coefficients for the transient processes after ECRH switch-on/off in the T-10 tokamak

International Nuclear Information System (INIS)

Andreev, V.F.; Danilov, A.V.; Dnestrovskij, Yu.N.; Ossipenko, M.V.; Razumova, K.A.; Sushkov, A.V.

2005-01-01

A study of the transient processes after ECRH switch-on/off in the T-10 tokamak shows that an electron ITB appears near the q∼3/2 resonance surface in regimes with suppressed sawtooth oscillations. When the gradient of the electron temperature in the plasma centre after additional on-axis ECRH exceeds some critical value, the electron ITB is weakened and the heat flux spreads from the heating region to the plasma periphery practically instantly. This looks like a jump of the heat flux through the entire plasma region. Therefore, to describe the electron heat transport in a tokamak plasma after the ECRH switch-on/off, three different physical processes should be taken into account: a) the transient process of plasma heating inside the region bounded by the temporal electron ITB; b) the weakening of the electron ITB accompanied by the heat efflux to the plasma periphery during hundreds of microseconds; c) the consequent diffusive evolution of the electron temperature profile with the conservation of the relative electron temperature gradient ∇T/T. (author)

Density profiles and particle fluxes of heavy impurities in the limiter shadow region of a tokamak

International Nuclear Information System (INIS)

Claassen, H.A.; Repp, H.

1980-01-01

For the case of low impurity concentration, transport calculations have been performed for heavy impurities, in the scrape-off layer plasma of a tokamak with a poloidal ring limiter. The theory is based on the drift-kinetic equations for the various ionization states of the impurity ions taking due consideration of the convection and collision processes. The background plasma and the impurity sources from the torus wall and the limiter surface enter the theory as input parameters. The theory is developed for the first two orders of the drift approximation. Numerical results are given to zero order drift approximation for the radial profiles of density and particle fluxes parallel to the magnetic field. (orig.)

Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

Science.gov (United States)

Chen, B.; Xu, X. Q.; Xia, T. Y.; Li, N. M.; Porkolab, M.; Edlund, E.; LaBombard, B.; Terry, J.; Hughes, J. W.; Ye, M. Y.; Wan, Y. X.

2018-05-01

The heat flux distributions on divertor targets in H-mode plasmas are serious concerns for future devices. We seek to simulate the tokamak boundary plasma turbulence and heat transport in the edge localized mode-suppressed regimes. The improved BOUT++ model shows that not only Ip but also the radial electric field Er plays an important role on the turbulence behavior and sets the heat flux width. Instead of calculating Er from the pressure gradient term (diamagnetic Er), it is calculated from the plasma transport equations with the sheath potential in the scrape-off layer and the plasma density and temperature profiles inside the separatrix from the experiment. The simulation results with the new Er model have better agreement with the experiment than using the diamagnetic Er model: (1) The electromagnetic turbulence in enhanced Dα H-mode shows the characteristics of quasi-coherent modes (QCMs) and broadband turbulence. The mode spectra are in agreement with the phase contrast imaging data and almost has no change in comparison to the cases which use the diamagnetic Er model; (2) the self-consistent boundary Er is needed for the turbulence simulations to get the consistent heat flux width with the experiment; (3) the frequencies of the QCMs are proportional to Er, while the divertor heat flux widths are inversely proportional to Er; and (4) the BOUT++ turbulence simulations yield a similar heat flux width to the experimental Eich scaling law and the prediction from the Goldston heuristic drift model.

The primary results for the mixed carbon material used for high flux steady-state tokamak operation in China

International Nuclear Information System (INIS)

Guo, Q.G.; Li, J.G.; Zhai, G.T.; Liu, L.; Song, J.R.; Zhang, L.F.; He, Y.X.; Chen, J.L.

2001-01-01

Several types of carbon mixed materials have been developed in China to be used for high flux steady-state tokamak operation. Performance evaluation of these materials is necessary to determine their applicability as PFCs for high flux steady state. This paper describes the primary results of carbon mixed materials and the effects of dopants on properties are primarily discussed. Test results reveal that bulk boronized graphite has excellent physical and mechanical properties while their thermal conductivity is no more than 73 W/m K due to the formation of a uniform boron-carbon solid solution. In case of multi-element doped graphite, titanium dopant or a decreased boron content is favorable to enhance thermal conductivity. A kind of doped graphite has been developed with thermal conductivity as high as 278 W/m K by optimizing the compositions. Correlations among compositions, microstructure and properties of such doped graphite are discussed

Total magnetic reconnection during a tokamak major disruption

International Nuclear Information System (INIS)

Goetz, J.A.

1990-09-01

Magnetic reconnection has long been considered to be the cause of sawtooth oscillations and major disruptions in tokamak experiments. Experimental confirmation of reconnection models has been hampered by the difficulty of direct measurement of reconnection, which would involve tracing field lines for many transits around the tokamak. Perhaps the most stringent test of reconnection in a tokamak involves measurement of the safety factor q. Reconnection arising from a single helical disturbance with mode numbers m and n should raise q to m/n everywhere inside of the original resonant surface. Total reconnection should also flatten the temperature and current density profiles inside of this surface. Disruptive instabilities have been studied in the Tokapole 2, a poloidal divertor tokamak. When Tokapole 2 is operated in the material limiter configuration, a major disruption results in current termination as in most tokamaks. However, when operated in the magnetic limiter configuration current termination is suppressed and major disruptions appear as giant sawtooth oscillations. The objective of this thesis is to determine if total reconnection is occurring during major disruptions. To accomplish this goal, the poloidal magnetic field has been directly measured in Tokapole 2 with internal magnetic coils. A full two-dimensional measurement over the central current channel has been done. From these measurements, the poloidal magnetic flux function is obtained and the magnetic surfaces are plotted. The flux-surface-averaged safety factor is obtained by integrating the local magnetic field line pitch over the experimentally obtained magnetic surface

Surface temperature measurement of plasma facing components in tokamaks

International Nuclear Information System (INIS)

Amiel, Stephane

2014-01-01

During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author) [fr

Erosion of the first wall of Tokamaks

International Nuclear Information System (INIS)

Guseva, M.I.; Ionova, E.S.; Martynenko, Yu.V.

1980-01-01

An estimate of the rate of erosion of the wall due to sputtering and blistering requires knowledge of the fluxes and energies of the particles which go from the plasma to the wall, of the sputtering coefficients S, and of the erosion coefficients S* for blistering. The overall erosion coefficient is equal to the sum of the sputtering coefficient and the erosion coefficient for blistering. Here the T-20 Tokamak is examined as an example of a large-scale Tokamak. 18 refs

Improvement of tokamak performance by injection of electrons

International Nuclear Information System (INIS)

Ono, Masayuki.

1992-12-01

Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas

Neutronic analysis of fusion tokamak devices by PHITS

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

2011-01-01

A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

Compression experiments on the TOSKA tokamak

International Nuclear Information System (INIS)

Cima, G.; McGuire, K.M.; Robinson, D.C.; Wootton, A.J.

1980-10-01

Results from minor radius compression experiments on a tokamak plasma in TOSCA are reported. The compression is achieved by increasing the toroidal field up to twice its initial value in 200μs. Measurements show that particles and magnetic flux are conserved. When the initial energy confinement time is comparable with the compression time, energy gains are greater than for an adiabatic change of state. The total beta value increases. Central beta values approximately 3% are measured when a small major radius compression is superimposed on a minor radius compression. Magnetic field fluctuations are affected: both the amplitude and period decrease. Starting from low energy confinement times, approximately 200μs, increases in confinement times up to approximately 1 ms are measured. The increase in plasma energy results from a large reduction in the power losses during the compression. When the initial energy confinement time is much longer than the compression time, the parameter changes are those expected for an adiabatic change of state. (author)

Anomalous transport in tokamaks

International Nuclear Information System (INIS)

Wootton, A.J.

1989-01-01

A review is presented of what is known about anomalous transport in tokamaks. It is generally thought that this anomalous transport is the result of fluctuations in various plasma parameters. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation driven fluxes are available. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge energy flux are less clear, but electrostatic convection plays an important part. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made. Magnetic fluctuations is discussed

Two-ion hybrid resonances and ion cyclotron absorption in tokamak plasmas

International Nuclear Information System (INIS)

Brambilla, M.; Ottaviani, M.

1983-11-01

The behaviour of IC waves near resonances in tokamak geometry is investigated in details. For this purpose, a one-dimensional model is proposed, which takes into account the orientation of the incident wavefronts with respect both to the singular layer and to the magnetic surfaces. The differential equations describing the waves are derived again from Vlasov-Maxwell equations in the finite Larmor radius approximation; they are shown to conserve the wave power flux in the absence of dissipation, and to reproduce the local dispersion relation in the WKB limit. These equations are solved exactly in some important situations, and with the Green-function technique in the general case. The amount of power coupled to Bernstein waves and absorbed by cyclotron damping is explicitly evaluated. (orig.)

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

High heat flux testing of CFC composites for the tokamak physics experiment

Science.gov (United States)

Valentine, P. G.; Nygren, R. E.; Burns, R. W.; Rocket, P. D.; Colleraine, A. P.; Lederich, R. J.; Bradley, J. T.

1996-10-01

High heat flux (HHF) testing of carbon fiber reinforced carbon composites (CFC's) was conducted under the General Atomics program to develop plasma-facing components (PFC's) for Princeton Plasma Physics Laboratory's tokamak physics experiment (TPX). As part of the process of selecting TPX CFC materials, a series of HHF tests were conducted with the 30 kW electron beam test system (EBTS) facility at Sandia National Laboratories, and with the plasma disruption simulator I (PLADIS-I) facility at the University of New Mexico. The purpose of the tests was to make assessments of the thermal performance and erosion behavior of CFC materials. Tests were conducted with 42 different CFC materials. In general, the CFC materials withstood the rapid thermal pulse environments without fracturing, delaminating, or degrading in a non-uniform manner; significant differences in thermal performance, erosion behavior, vapor evolution, etc. were observed and preliminary findings are presented below. The CFC's exposed to the hydrogen plasma pulses in PLADIS-I exhibited greater erosion rates than the CFC materials exposed to the electron-beam pulses in EBTS. The results obtained support the continued consideration of a variety of CFC composites for TPX PFC components.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Expanded-boundary approach to impurity control in tokamaks

International Nuclear Information System (INIS)

Ohyabu, N.

1981-01-01

It is proposed to expand the outermost flux surfaces in tokamaks to divert the heat flux emerging from the plasma core. The expanded flux surfaces provide a large volume for radiative cooling. The radiative power at the boundary is enhanced by the effects of plasma flow as well as by a volumetric factor, and the resultant edge cooling and reduced heat load on the limiter may significantly retard impurity generation. Furthermore, it seems to be compatible with reactor engineering requirements. (author)

Configuration of gun-generated spheromak in effectively closed metal flux conserver

International Nuclear Information System (INIS)

Kato, Yushi; Nishikawa, Masahiro; Honda, Yoshihide; Satomi, Norio; Watanabe, Kenji

1988-01-01

In the CTCC-II spheromak experiment, the gun-generated plasma is confined in a spheroidal aluminum flux conserver (FC) with a choking coil. This coil produces the additional magnetic field to close perfectly all magnetic surfaces into the FC, i.e. the entrance hole for plasma injection is enable to be closed by magnetic field. Hence the plasma is confined in the effectively closed metal FC. In this experiment the average life time is 1.2 msec, and electron density and temperature are n e = 2 x 10 13 /cc, T e = 30 eV, respectively. The configuration with a flux hole region in which the toroidal magnetic field vanishes around the geometrical axis has been observed in the FC. The radius of the flux hole depends on the condition how the external choking field is applied. The flux hole enhances the magnetic shear near the plasma surfaces and, therefore, has a stabilizing effect even without inserting the central conducting pole. (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)

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

Realizing steady-state tokamak operation for fusion energy

International Nuclear Information System (INIS)

Luce, T. C.

2011-01-01

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

Gamma ray imager on the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.; Van Zeeland, M. A.; Watkins, M. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Cooper, C. M. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Riso, V. [State University of New York-Buffalo, 12 Capen Hall, Buffalo, New York 14260-1660 (United States)

2016-04-15

A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1–60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

Stability of high β large aspect ratio tokamaks

International Nuclear Information System (INIS)

Cowley, S.C.

1991-10-01

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

Scrape-off layer flows in the Tore Supra tokamak

International Nuclear Information System (INIS)

Gunn, J.P.; Loarer, T.; Saint-Laurent, F.; Bucalossi, J.; Devynck, P.; Hertout, P.; Moreau, P.; Nanobashvili, I.; Rimini, F.; Duran, I.; Fuchs, V.; Panek, R.; Stockel, J.; Adamek, J.; Dejarnac, R.; Hron, M.; Sarkissian, A.

2005-01-01

Near-sonic parallel flows are systematically observed in the scrape-off layer (SOL) of the limiter tokamak Tore Supra, as in many X-point divertor tokamaks. The poloidal variation of the Mach number of the parallel flow has been measured by moving the contact point of a small circular plasma onto limiters at different poloidal angles. The resulting variations of flow are consistent with the existence of a poloidally nonuniform core-to-SOL out-flux concentrated near the outboard midplane. Strong variations of the SOL width up to a factor of 10 suggest that this localized out-flux is due to enhanced radial transport. The plasma that gets ejected into the SOL can expand radially to the wall if magnetic field lines have long connection lengths and pass unobstructed across the outboard midplane. (authors)

Calculations of axisymmetric stability of tokamak plasmas with active and passive feedback

International Nuclear Information System (INIS)

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

1991-07-01

A new linear MHD stability code, NOVA-W, has been developed in order to study feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The passive stability predictions of the code have been tested both against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops that provide the measurements of vertical displacement. The results compare well with those computed earlier using a less efficient nonlinear code. 37 refs., 13 figs

Simulation of tokamak armour erosion and plasma contamination at intense transient heat fluxes in ITER

Science.gov (United States)

Landman, I. S.; Bazylev, B. N.; Garkusha, I. E.; Loarte, A.; Pestchanyi, S. E.; Safronov, V. M.

2005-03-01

For ITER, the potential material damage of plasma facing tungsten-, CFC-, or beryllium components during transient processes such as ELMs or mitigated disruptions are simulated numerically using the MHD code FOREV-2D and the melt motion code MEMOS-1.5D for a heat deposition in the range of 0.5-3 MJ/m 2 on the time scale of 0.1-1 ms. Such loads can cause significant evaporation at the target surface and a contamination of the SOL by the ions of evaporated material. Results are presented on carbon plasma dynamics in toroidal geometry and on radiation fluxes from the SOL carbon ions obtained with FOREV-2D. The validation of MEMOS-1.5D against the plasma gun tokamak simulators MK-200UG and QSPA-Kh50, based on the tungsten melting threshold, is described. Simulations with MEMOS-1.5D for a beryllium first wall that provide important details about the melt motion dynamics and typical features of the damage are reported.

Particle and heat balance analysis in scrape-off and divertor regions of the JFT-2M tokamak

International Nuclear Information System (INIS)

Nagashima, K.; Shoji, T.; Tamai, H.; Miura, Y.; Takenaga, H.; Maeda, H.

1995-01-01

Particle and heat balance in the scrape-off layer and the divertor region were studied in the JFT-2M tokamak. Using particle and energy conservation laws, particle and heat diffusivities perpendicular to the flux surface were evaluated just outside the magnetic separatrix. It was found that the particle diffusivity decreases with increasing electron density in the scrape-off layer and decreases by a factor of 2-3 in the H-mode phase as compared with that in L-mode. The heat diffusivity has almost the same dependence on the electron density. The ratio of the heat diffusivity to the particle diffusivity is about 2. ((orig.))

Physics objectives of PI3 spherical tokamak program

Science.gov (United States)

Howard, Stephen; Laberge, Michel; Reynolds, Meritt; O'Shea, Peter; Ivanov, Russ; Young, William; Carle, Patrick; Froese, Aaron; Epp, Kelly

2017-10-01

Achieving net energy gain with a Magnetized Target Fusion (MTF) system requires the initial plasma state to satisfy a set of performance goals, such as particle inventory (1021 ions), sufficient magnetic flux (0.3 Wb) to confine the plasma without MHD instability, and initial energy confinement time several times longer than the compression time. General Fusion (GF) is now constructing Plasma Injector 3 (PI3) to explore the physics of reactor-scale plasmas. Energy considerations lead us to design around an initial state of Rvessel = 1 m. PI3 will use fast coaxial helicity injection via a Marshall gun to create a spherical tokamak plasma, with no additional heating. MTF requires solenoid-free startup with no vertical field coils, and will rely on flux conservation by a metal wall. PI3 is 5x larger than SPECTOR so is expected to yield magnetic lifetime increase of 25x, while peak temperature of PI3 is expected to be similar (400-500 eV) Physics investigations will study MHD activity and the resistive and convective evolution of current, temperature and density profiles. We seek to understand the confinement physics, radiative loss, thermal and particle transport, recycling and edge physics of PI3.

Tokamak power plant burn cycle options

International Nuclear Information System (INIS)

Ehst, D.A.

1994-06-01

Experiments show that tokamaks can operate in various fashions. Economic analyses show that steady state is most attractive provided the physics and technology of current drive (CD) can be modestly improved. Even with very conservative CD assumptions a hybrid operating mode seems superior to conventional, simple inductive operation

Neoclassical MHD equations for tokamaks

International Nuclear Information System (INIS)

Callen, J.D.; Shaing, K.C.

1986-03-01

The moment equation approach to neoclassical-type processes is used to derive the flows, currents and resistive MHD-like equations for studying equilibria and instabilities in axisymmetric tokamak plasmas operating in the banana-plateau collisionality regime (ν* approx. 1). The resultant ''neoclassical MHD'' equations differ from the usual reduced equations of resistive MHD primarily by the addition of the important viscous relaxation effects within a magnetic flux surface. The primary effects of the parallel (poloidal) viscous relaxation are: (1) Rapid (approx. ν/sub i/) damping of the poloidal ion flow so the residual flow is only toroidal; (2) addition of the bootstrap current contribution to Ohm's laws; and (3) an enhanced (by B 2 /B/sub theta/ 2 ) polarization drift type term and consequent enhancement of the perpendicular dielectric constant due to parallel flow inertia, which causes the equations to depend only on the poloidal magnetic field B/sub theta/. Gyroviscosity (or diamagnetic vfiscosity) effects are included to properly treat the diamagnetic flow effects. The nonlinear form of the neoclassical MHD equations is derived and shown to satisfy an energy conservation equation with dissipation arising from Joule and poloidal viscous heating, and transport due to classical and neoclassical diffusion

Is X-ray emissivity constant on magnetic flux surfaces?

International Nuclear Information System (INIS)

Granetz, R.S.; Borras, M.C.

1997-01-01

Knowledge of the elongations and shifts of internal magnetic flux surfaces can be used to determine the q profile in elongated tokamak plasmas. X-ray tomography is thought to be a reasonable technique for independently measuring internal flux surface shapes, because it is widely believed that X-ray emissivity should be constant on a magnetic flux surface. In the Alcator C-Mod tokamak, the X-ray tomography diagnostic system consists of four arrays of 38 chords each. A comparison of reconstructed X-ray contours with magnetic flux surfaces shows a small but consistent discrepancy in the radial profile of elongation. Numerous computational tests have been performed to verify these findings, including tests of the sensitivity to calibration and viewing geometry errors, the accuracy of the tomography reconstruction algorithms, and other subtler effects. We conclude that the discrepancy between the X-ray contours and the magnetic flux surfaces is real, leading to the conclusion that X-ray emissivity is not exactly constant on a flux surface. (orig.)

Transport in the tokamak plasma edge

International Nuclear Information System (INIS)

Vold, E.L.

1989-01-01

Experimental observations characterize the edge plasma or boundary layer in magnetically confined plasmas as a region of great complexity. Evidence suggests the edge physics plays a key role in plasma confinement although the mechanism remains unresolved. This study focuses on issues in two areas: observed poloidal asymmetries in the Scrape Off Layer (SOL) edge plasma and the physical nature of the plasma-neutral recycling. A computational model solves the coupled two dimensional partial differential equations governing the plasma fluid density, parallel and radial velocities, electron and ion temperatures and neutral density under assumptions of toroidal symmetry, ambipolarity, anomalous diffusive radial flux, and neutral-ion thermal equilibrium. Drift flow and plasma potential are calculated as dependent quantities. Computational results are compared to experimental data for the CCT and TEXTOR:ALT-II tokamak limiter cases. Comparisons show drift flux is a major component of the poloidal flow in the SOL along the tangency/separatrix. Plasma-neutral recycling is characterized in several tokamak divertors, including the C-MOD device using magnetic flux surface coordinates. Recycling is characterized by time constant, τ rc , on the order of tens of milliseconds. Heat flux transients from the core into the edge on shorter time scales significantly increase the plasma temperatures at the target and may increase sputtering. Recycling conditions in divertors vary considerably depending on recycled flux to the core. The high density, low temperature solution requires that the neutral mean free path be small compared to the divertor target to x-point distance. The simulations and analysis support H-mode confinement and transition models based on the recycling divertor solution bifurcation

Searching for Conservation Laws in Brain Dynamics—BOLD Flux and Source Imaging

Directory of Open Access Journals (Sweden)

Henning U. Voss

2014-07-01

Full Text Available Blood-oxygen-level-dependent (BOLD imaging is the most important noninvasive tool to map human brain function. It relies on local blood-flow changes controlled by neurovascular coupling effects, usually in response to some cognitive or perceptual task. In this contribution we ask if the spatiotemporal dynamics of the BOLD signal can be modeled by a conservation law. In analogy to the description of physical laws, which often can be derived from some underlying conservation law, identification of conservation laws in the brain could lead to new models for the functional organization of the brain. Our model is independent of the nature of the conservation law, but we discuss possible hints and motivations for conservation laws. For example, globally limited blood supply and local competition between brain regions for blood might restrict the large scale BOLD signal in certain ways that could be observable. One proposed selective pressure for the evolution of such conservation laws is the closed volume of the skull limiting the expansion of brain tissue by increases in blood volume. These ideas are demonstrated on a mental motor imagery fMRI experiment, in which functional brain activation was mapped in a group of volunteers imagining themselves swimming. In order to search for local conservation laws during this complex cognitive process, we derived maps of quantities resulting from spatial interaction of the BOLD amplitudes. Specifically, we mapped fluxes and sources of the BOLD signal, terms that would appear in a description by a continuity equation. Whereas we cannot present final answers with the particular analysis of this particular experiment, some results seem to be non-trivial. For example, we found that during task the group BOLD flux covered more widespread regions than identified by conventional BOLD mapping and was always increasing during task. It is our hope that these results motivate more work towards the search for conservation

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)

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-03-01

This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

Theory-based scaling of the SOL width in circular limited tokamak plasmas

International Nuclear Information System (INIS)

Halpern, F.D.; Ricci, P.; Labit, B.; Furno, I.; Jolliet, S.; Loizu, J.; Mosetto, A.; Arnoux, G.; Silva, C.; Gunn, J.P.; Horacek, J.; Kočan, M.; LaBombard, B.

2013-01-01

A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement. (letter)

Alpha transport and blistering in tokamaks

International Nuclear Information System (INIS)

Bauer, W.; Wilson, K.L.; Bisson, C.L.; Haggmark, L.G.; Goldston, R.J.

1978-12-01

The particle flux and angular distribution of 3.5 MeV alpha particles impinging on the first wall from uncontained banana orbits in an axisymmetric tokamak reactor have been calculated. The resulting helium concentration profiles in the first wall can give rise to surface exfoliation under specified conditions. The major mitigating factor is the simultaneous surface recession due to sputtering by the D-T charge exchange neutral flux. For the parameters used in these calculations blistering in high sputtering rate materials such as Be is unlikely whereas in low sputtering rate materials such as Nb, He induced surface deformation is quite probable

Physics of tokamak scrape-off layer confinement

International Nuclear Information System (INIS)

Cohen, R.H.

1993-01-01

Confinement in the scrape-off layer (SOL) of a tokamak is believed to be governed by classical flows along magnetic field lines terminated by sheaths, and turbulent transport across field lines. In this paper we review how these two effects conspire to establish the width of the SOL, and survey recent and ongoing work on mechanisms for turbulence in SOL's. The beneficial relationship between scrape-off layer turbulence in mitigating the heat flux density on divertors is noted, and tactics for actively altering SOL confinement so as to reduce the heat flux density are discussed

The calculation of Tritium burnup in Tokamaks

International Nuclear Information System (INIS)

Bittoni, E.; Haegi, M.

1987-01-01

In a deuterium plasma tokamak, the contained fusion-produced tritons are supposed to be decelerated down to thermalization according to classical Coulomb scattering. A fraction of these fast tritons undergoes the DT fusion reaction producing 14.1 MeV neutrons. It is thus possible to get information on the confinement of these fast tritons by comparing the measured and the calculated ratio of the 14.1 MeV to the 2.45 MeV neutron flux. This report describes the calculation of this flux ratio by means of a numerical Monte Carlo-like code

The simple map for a single-null divertor tokamak

International Nuclear Information System (INIS)

Punjabi, A.; Verma, A.; Boozer, A.

1996-01-01

We present the simple map for a single-null divertor tokamak. The simple map is an area-preserving map based on the idea that magnetic field lines are a single-degree-of-freedom time-dependent Hamiltonian system, and that the basic features of such systems near the X-point are generic. We obtain the properties of this map and the resulting footprints of field lines on the divertor plate. These include the width of the stochastic layer, the edge safety factor, the area of the footprint and the amount of magnetic flux diverted. We give the safety factor profile, the average and median values of strike angles, lengths and the Liapunov exponents. We describe how the effects of magnetic perturbations can be included in the simple map. We show how the map can be applied to the problem of the determination of heat flux on the divertor plate in tokamaks. (Author)

An investigation on detection and measurement of fusion neutron spectrum and radiation flux in large tokamak

International Nuclear Information System (INIS)

Yang Jinwei; Li Wenzhong; Zhang Wei

2003-01-01

The detection methods, detectors and spectrometers of D-D and D-T fusion neutron have been overviewed in large tokamaks. Some options are proposed for developing new detection systems of fusion neutrons suitable to the HL-2A tokamak. (authors)

The critical temperature gradient model of plasma transport: applications to Jet and future tokamaks

International Nuclear Information System (INIS)

Rebut, P.H.; Lallia, P.P.; Watkins, M.L.

1989-01-01

The diversity and complexity of behaviour in tokamak plasmas place strong constraints on any model attempting a description in terms of a single underlying phenomenon. Assuming that turbulence in the magnetic topology is the underlying phenomenon, specific expressions for electron and ion heat flux are derived from heuristic and dimensional arguments. When used in plasma transport codes, rather satisfactory simulations of experimental results are achieved in different sized tokamaks in various regimes of operation. Predictions are given for the expected performance of JET at full planned power and implications for next step tokamaks are indicated

Issues for the electric utilities posed by DT tokamak fusion powerplants

International Nuclear Information System (INIS)

Roth, J.R.

1990-01-01

The DT tokamak is the mainline approach to magnetic fusion energy in all industrialized countries with a major commitment to fusion research. It achieved this status largely through historical accident and not as the result of considered choice among alternatives. After twenty-five years of intensive tokamak research, it is appropriate to ask whether the path down which the tokamak concept is leading the fusion community is the way to an acceptable powerplant for the electric utilities, or an aberration which should be replaced with an approach more promising in the long term. Issues surrounding the DT tokamak can be grouped in three broad areas: physics; safety/environmental; and engineering/economic. In addition to these problems, detailed engineering design studies of DT tokamak fusion powerplants over a twenty year period have revealed a number of additional problems. Most of thee are related to the presence of tritium and energetic neutron fluxes, which tend to make the cost of electricity of DT tokamaks higher than that of fossil or fission powerplants. These safety and economic issues of the DT tokamak powerplant also appear to be intractable, and have not been made to go away by twenty years of progressively more detailed and extensive engineering design studies

Experimental study on the practicability of a self-created spherical tokamak in the coil less STPC-EX machine

International Nuclear Information System (INIS)

Sinman, S.; Sinman, A.

2003-01-01

The aim of this study is to identify the physical basis of the alternative self-organization mechanism that exists on the STPC-EX machine and to determine complementary features with respect to present compact toroid concepts. In the STPC-EX machine, there exist two solenoids placed inside the central passive floating conductive hollow rod and externally onto flux conserver. These are in a passive state and they do not have an important role in the self-created spherical tokamak plasma (SCSTP) in the STPC-EX machine. In this study, conventional diagnostic tools are used and for photographic recording, the method of open shutter integrated post-fogging is chosen. Two annular coaxial plasma current sheets, one within the other in the same direction, are created and flow on the surface of the central conductive hollow rod. Consequently, the spherical tokamak is configured by a new creation mechanism called the dual-axial z-pinch. High betas of 0.4-0.6 and aspect ratios of up to 1.25 can be obtained. (author)

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)

Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR

Energy Technology Data Exchange (ETDEWEB)

Sa, Wanderley Pires de

1996-12-31

The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.

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

On 2X2 systems of conservation laws with fluxes that are entropies

Directory of Open Access Journals (Sweden)

Michael Junk

2003-03-01

Full Text Available In this article, we study systems of conservation laws with two dependent and two independent variables which have the property that the fluxes are entropies. Several characterizations of such flux functions are presented. It turns out, that the corresponding systems automatically possess a large class of additional entropies, they are closely related to a kinetic equation, and, in the case of strict hyperbolicity, they can be decoupled into two independent Burgers' equations. The isentropic Euler equations with zero or cubic pressure laws are the most prominent examples of such systems, but other examples are also presented.

Low Z impurity transport in tokamaks

International Nuclear Information System (INIS)

Hawryluk, R.J.; Suckewer, S.; Hirshman, S.P.

1978-10-01

Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed

Tokamak

International Nuclear Information System (INIS)

Wesson, John.

1996-01-01

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

Intrinsic non-inductive current driven by ETG turbulence in tokamaks

Science.gov (United States)

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

2017-10-01

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

Current diffusion and flux consumption in Tore Supra

International Nuclear Information System (INIS)

Van Houtte, D.; Talvard, M.; Agostini, E.; Gil, C.; Hoang, G.T.; Lecoustey, P.; Parlange, F.; Rodriguez, L.; Vallet, J.C.

1991-01-01

TORE SUPRA has been designed to study long pulse plasmas (t > 30 s) at high plasma current (Ip < 2 MA) associated with high additional power (20 MW). Current diffusion studies are essentially based on the analysis of the plasma discharge paths. The current diffusion rate during the current rise phase is analysed with a numerical code using plasma resistivity profiles from Te profiles measured by the ECE diagnostic. Owing to the fact that the quantity of magnetic flux available in a tokamak is limited, perfect knowledge is required of the various components of the flux consumed in order to minimize consumption and to be able to define a suitable transformer size for future high current tokamak projects

Rapidly Moving Divertor Plates In A Tokamak

International Nuclear Information System (INIS)

Zweben, S.

2011-01-01

It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

Studies on fundamental technologies for producing tokamak-plasma

International Nuclear Information System (INIS)

Matsuzaki, Yoshimi

1987-10-01

The report describes studies on fundamental technologies to produce tokamak-plasma of the JFT-2 and JFT-2M tokamaks. (1) In order to measure the particle number of residual gases, calibration methods of vacuum gauges have been developed. (2) Devices for a Taylor-type discharge cleaning (TDC), a glow discharge cleaning (GDC) and ECR discharge cleaning (ECR-DC) have been made and the cleaning effects have been investigated. In TDC the most effective plasma for cleaning is obtained in the plasma with 5 eV of electron temperature. GDC is effective in removing carbon impurities, but is less effective for removing oxygen impurities. ECR-DC has nearly the similar effect as TDC. The cleaning effect of these three types were studied by comparing the properties of resulting tokamak plasmas in the JFT-2M tokamak. (3) Experimental studies of pre-ionization showed as following results; A simple pre-ionization equipment as a hot-electron-gun and a J x B gun was effective in reducing breakdown voltage. An ordinary mode wave of the electron cyclotron frequency was very effective for pre-ionization. The RF power whose density is 3.6 x 10 -2 W/cm 3 produced plasma of an electron density of 5 x 10 11 cm -3 . In this case, it is possible to start up with negligible consumption of the magnetic flux caused by the plasma resistance. (4) Concerning to studies on plasma control, the following results were obtained; In order to obtain constant plasma current, a pulse forming network was constructed and sufficient constant plasma current was achieved. In applying an iso-flux method for measuring the plasma position, it is no problem practically to use only one loop-coil and one magnetic probe. (author)

Calculation of the neoclassical conductivity of plasma and fraction of trapped particles for elongated Damavand Tokamak

International Nuclear Information System (INIS)

Dini, F.; Khorasani, S.

2007-01-01

Configuration of Tokamak plasma has a dominant effect on its parameters. In the calculation of transport, there are some transport coefficients and quantities, where the knowledge of their precise values, according to the system of equations, is essential to be realized. Tokamak has a toroidal configuration, in addition to classical effects, it is necessary to study the neoclassical effects due to the field curvature. The trapped particles in strong electromagnetic fields oscillate on banana-shaped orbits which in turn affect many other collisional transport parameters. Here, a precise estimation of trapped particles based on the standard equilibrium model for an elliptical shape of Tokamak plasma has been carried out using Lin-Liu model. It should be added that in this calculation, the profile of the averaged magnetic field on the flux surfaces has been derived using analytical integration and consideration of an elliptic shape for ellipticity function in the limit of large aspect ratio and zero shift of magnetic flux surfaces. Having the fraction of the trapped particles, by ,following the formulation and using an appropriate model in various collisional regimes, the neoclassical conductivity of plasma in Damavand Tokamak is obtained and the respective variations have been found. The presented results can exploit the computation of transport and other quantities of Damavand Tokamak

Developing physics basis for the snowflake divertor in the DIII-D tokamak

Science.gov (United States)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

Theory of ion heat transport in tokamaks

International Nuclear Information System (INIS)

Gott, Y.V.; Yurchenko, E.I.

1987-01-01

Experiments which have been carried out in several tokamaks to determine the ion thermal conductivity show that it is several times the value predicted by the neoclassical theory. A possible explanation for this discrepancy is proposed. When the finite width of a banana is taken into account, there are substantial increases in the heat fluxes which stem from the important contribution of superthermal ions to the transport. If the electron diffusive flux is zero, a systematic account of the ions with E>T leads to an ion heat flux with a finite banana width which is two to four times the neoclassical prediction. The effect of the anomalous nature of the electron flux on the ion heat transport is analyzed. An expression is derived for calculating the ion heat transport over the entire range of collision rates

Feedback control of edge turbulence in a tokamak

International Nuclear Information System (INIS)

Kan, Zhai; Yi-zhi, Wen; Chang-xuan, Yu; Wan-dong, Liu; Chao, Wang; Ge, Zhuang; Kan, Zhai; Zhi-Zhan, Yu

1997-01-01

An experiment on feedback control of edge turbulence has been undertaken on the KT-5C tokamak. The results indicate that the edge turbulence could be suppressed or enhanced depending on the phase shift of the feedback network. In a typical case of 90 degree phase shift feedback, the turbulence amplitudes of both T e and n e were reduced by about 25% when the gain of the feedback network was 15. Correspondingly the radial particle flux decreased to about 75% level of the background. Through bispectral analysis it is found that there exists a substantial nonlinear coupling between various modes comprised in edge turbulence, especially in the frequency range from about 10 kHz to 100 kHz, which contains the large part of the edge turbulence energy in KT-5C tokamak. In particular, by actively controlling the turbulence amplitude using feedback, a direct experimental evidence of the link between the nonlinear wave-wave coupling over the whole spectrum in turbulence, the saturated turbulence amplitude, and the radial particle flux was provided. copyright 1997 The American Physical Society

Quantitative spectroscopy of X-ray lines and continua in tokamaks

International Nuclear Information System (INIS)

Peacock, N.J.; Lawson, K.D.; Patel, A.; Barnsley, R.; Melnick, I.M.; O'Mullane, M.G.; Singleton, M.A.

1996-10-01

Crystal and synthetic multilayer diffractors, deployed either as flat Bragg reflectors, or curved, as in the Johann configuration, are used to study the spectrum of COMPASS-D and other tokamaks in the wavelength region 1-100 A. In this paper we concentrate on the measurement of absolute photon fluxes and the derivation of volume emissivities of the lines and continua in the X-ray region. The sensitivities of these instruments to absolute photon flux have been constructed ab initio from the individual component efficiencies, including published values of the diffractor reflectivities, which have been checked or supplemented by measurements using a double-axis goniometer or from line branching ratios. For these tokamak plasmas where the elemental abundances and effective ion charge are documented, the X-ray continuum intensity itself has been used as a calibration source to derive absolute instrument sensitivity, in reasonable agreement with the ab initio method. In the COMPASS-D tokamak, changes in the effective ion charge state, Z eff , have been derived for different operating conditions, from the absolute intensity of the continuum at ∼4 A. From the irradiances of the line emission, changes in the absolute level of impurities following ''boronisation'' of the vacuum vessel have also been documented. (UK)

Thick boron carbide coatings for protection of tokamak first wall and divertor

International Nuclear Information System (INIS)

Buzhinskij, O.I.; Semenets, Yu.M.

1999-01-01

A review of characteristics of various types of boron carbide coatings considered as candidate materials for protection of tokamak inner surfaces against high energy heat fluxes is presented. Such coatings are produced by various methods: chemical vapor deposition by means of chloride and fluoride techniques, gas conversion, plasma spray and reaction-sintering. Contrary to pure carbon materials, B 4 C has much lower chemical and high-temperature sputtering, is capable to oxygen gettering and lower hydrogen recycling. In contrast to thin boronization films, the thick coatings can resist high heat fluxes such as in tokamak divertors. Comparative analysis shows that coatings produced by the diffusion methods, such as fluoride CVD and gas conversion, are more resistent to heat loads, and one of the most promising candidates are the fluoride CVD coatings. (orig.)

Advanced divertor configurations with large flux expansion

Energy Technology Data Exchange (ETDEWEB)

Soukhanovskii, V.A., E-mail: vlad@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Bell, R.E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Menard, J.E.; Paul, S.F.; Podesta, M. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Raman, R. [University of Washington, Seattle, WA (United States); Ryutov, D.D. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Scotti, F.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mueller, D.M.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Reimerdes, H.; Canal, G.P. [Ecole Polytechnique Fédérale de Lausanne, Centre de Recherches en Physique des Plasmas, Association Euratom Confédération Suisse, Lausanne (Switzerland); and others

2013-07-15

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3–7 MW/m{sup 2} to 0.5–1 MW/m{sup 2} was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L–H power threshold, enhanced stability of the peeling–ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2–3) Type I ELM frequency and slightly increased (20–30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX

Runaway electrons during tokamak startup

International Nuclear Information System (INIS)

Sharma, A.S.; Jayakumar, R.

1988-01-01

Runaway electrons significantly affect the plasma and impurity evolution during tokamak startup. During its rise, a runaway pulse stores magnetic flux inductively; this is then released during the decay phase of the runaway pulse. This process affects plasma formation, current initiation and current buildup. Because of their relativistic velocities the runaway electrons have higher ionization and excitation rates than the plasma electrons. This leads to a significant modification of the impurity behaviour and consequently the plasma evolution. (author). 20 refs, 8 figs

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

Scrape-off measurements during Alfven wave heating in the TCA tokamak

International Nuclear Information System (INIS)

Hofmann, F.; Hollenstein, C.; Joye, B.; Lietti, A.; Lister, J.B.; Pochelon, A.; Gimzewski, J.K.; Veprek, S.

1984-01-01

Plasma parameters and impurity fluxes in the scrape-off layer of the TCA tokamak have been measured during Alfven wave heating. Langmuir probes are used to measure electron density, electron temperature and plasma potential. Collection probes, in conjunction with XPS surface analysis, are used to determine impurity fluxes and ion impact energies. During RF heating, the electron edge temperature rises, the plasma potential drops and impurity fluxes are enhanced. Probe erosion due to impurity sputtering is clearly observed. The measurements are correlated with other diagnostics on TCA. (orig.)

Drift wave turbulence studies on closed and open flux surfaces: effect limiter/divertor plates location

International Nuclear Information System (INIS)

Ribeiro, T.; Scott, B.

2007-01-01

The field line connection of a tokamak sheared magnetic field has an important impact on turbulence, by ensuring a finite parallel dynamical response for every degree of freedom available in the system. This constitutes the main property which distinguishes closed from open flux surfaces in such a device. In the latter case, the poloidal periodicity of the magnetic field is replaced by a Debye sheath arising where the field lines strike the limiter/divertor plates. This is enough to break the field line connection constraint and allow the existence of convective cell modes, leading to a change in the character of the turbulence from drift wave- (closed flux surfaces) to interchange-type (open flux surfaces), and hence increasing the turbulent transport observed. Here we study the effect of changing the poloidal position of the limiter/divertor plates, using the three-dimensional electromagnetic gyrofluid turbulence code GEM, which has time dependently self consistent field aligned flux tube coordinates. For the closed flux surfaces, the globally consistent periodic boundary conditions are invoked, and for open flux surfaces a standard Debye sheath is used at the striking points. In particular, the use of two limiter positions simultaneously, top and bottom, is in order, such to allow a separation between the inboard and outboard sides of the tokamak. This highlights the differences between those two regions of the tokamak, where the curvature is either favourable (former) or unfavourable (latter), and further makes room for future experimental qualitative comparisons, for instance, on double null configurations of the tokamak ASDEX Upgrade. (author)

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

Steady state operation of tokamaks. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2000-10-01

The first IAEA Technical Committee Meeting (TCM) on Steady State Operation of Tokamaks was organized to discuss the operations of present long-pulse tokamaks (TRIAM-1M, TORE SUPRA, MT-7, HT-7M, HL-1M) and the plans for future steady-state tokamaks such as SST-1, CIEL, and HT-7U. This meeting, held from 13-15 October 1998, was hosted by the Academia Sinica Institute of Plasma Physics (ASIPP), Hefei, China. Participants from China, France, India, Japan, the Russian Federation, and the IAEA participated in the meeting. There were 18 individual presentations plus general discussions on many topics, including superconducting magnet systems, cryogenics, plasma position control, non-inductive current drive, auxiliary heating, plasma-wall interactions, high heat flux components, particle control, and data acquisition

Scrape-off layer power flux measurements in the Tore Supra tokamak

Czech Academy of Sciences Publication Activity Database

Gunn, J. P.; Dejarnac, Renaud; Devynck, P.; Fedorczak, N.; Fuchs, Vladimír; Gil, C.; Kočan, M.; Komm, Michael; Kubič, M.; Lunt, T.; Monier-Garbet, P.; Pascal, J.-Y.; Saint-Laurent, F.

2013-01-01

Roč. 438, suppl (2013), S184-S188 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/20./. Aachen, 21.05.2012-25.05.2012] Institutional support: RVO:61389021 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.016, year: 2013 http://www.sciencedirect.com/science/article/pii/S0022311513000639#

Control of particles flux in a tokamak with an events structure; Controle des flux de particules dans un Tokamak au moyen d`une structure a events

Energy Technology Data Exchange (ETDEWEB)

Tsitrone, E

1995-12-01

Two key problems in the development of a controlled fusion reactor are: -the control of the ashes resulting from the fusion reaction (helium) and of the impurities coming from the wall erosion, which affect the central plasma performances by diluting the fuel and dissipating a part of the produced energy by radiation. - the removal of the heat carried to the walls by charged particles, which is highly concentrated (peak values of several tens of MW per m{sup 2}). Two types of systems are generally used for the plasma-wall interface: throat limiter and axisymmetric divertor. Neither is an ideal candidate to control simultaneously the heat and particle fluxes. This thesis investigates an alternative configuration, the vented limiter, tested for the first time on the Tore Supra tokamak. The vented limiter principle lies on the recycling neutrals collection by slots, in such a way that local thermal overload is avoided. It is shown experimentally that the surface temperature of the prototype installed in Tore Supra remains uniform. As far as the particle collection is concerned, even though the pressure in the vented limiter is lower than the pressure in the throat limiter by a factor 3 for deuterium and 4 helium, it is sufficient to control the plasma density. Moreover, as with a throat limiter, the pressure exhibits a quadratic evolution with the plasma density. To interpret these results, a model describing the plasma recycling on the limiter and the pumping by the slots has been developed. The model has been validated by a comparison with the experimental data. It was then used to propose an optimized version of the prototype with reshaped slots. This should improve the pumping efficiency by a factor 2, in deuterium as well as in helium, but without removing the discrepancy between both pumping efficiencies. (Abstract Truncated)

Viscosity in the edge of tokamak plasmas

International Nuclear Information System (INIS)

Stacey, W.M.

1993-05-01

A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the ''short-radial-gradient-scale-length'' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates

Modelling the turbulent transport of angular momentum in tokamak plasmas - A quasi-linear gyrokinetic approach

International Nuclear Information System (INIS)

Cottier, Pierre

2013-01-01

The magnetic confinement in tokamaks is for now the most advanced way towards energy production by nuclear fusion. Both theoretical and experimental studies showed that rotation generation can increase its performance by reducing the turbulent transport in tokamak plasmas. The rotation influence on the heat and particle fluxes is studied along with the angular momentum transport with the quasi-linear gyro-kinetic eigenvalue code QuaLiKiz. For this purpose, the QuaLiKiz code is modified in order to take the plasma rotation into account and compute the angular momentum flux. It is shown that QuaLiKiz framework is able to correctly predict the angular momentum flux including the E*B shear induced residual stress as well as the influence of rotation on the heat and particle fluxes. The major approximations of QuaLiKiz formalisms are reviewed, in particular the ballooning representation at its lowest order and the eigenfunctions calculated in the hydrodynamic limit. The construction of the quasi-linear fluxes is also reviewed in details and the quasi-linear angular momentum flux is derived. The different contributions to the turbulent momentum flux are studied and successfully compared both against non-linear gyro-kinetic simulations and experimental data. (author) [fr

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)

Investigation of a limiter scrape-off layer and its screening effect on LIBTOR tokamak

International Nuclear Information System (INIS)

Leonov, V.M.; Malakhov, N.P.; Alabyad, A.M.

1986-01-01

The main purpose of this work is to investigate screening properties of the limiter scrape-off layer (LSL), define the optimum width of this layer and investigate how energy and particle fluxes transfer from plasma to surrounding material surfaces from in the Libtor tokamak. For this, two moving rail molybdenum limiters (large as a working limiter and small as a probe), equipped with shunts and thermocouples were placed on the bottom side of tokamak at 980 deg apart the torus. The main results of the work are as follows: the introduction of a limiter allows an essential shielding of the chamber wall from the interaction with plasma and a reduction in the impurity influx into the plasma column. The main mechanism of the limiter erosion in the stationary part of a discharge is the ion sputtering. An active control over the limiter sputtering, over the heat and particle fluxes onto the limiter and over the plasma confinement time can be found to be of importance for the prospects of the limiter application as an alternative to the divertor in the tokamak-reactor

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Design of plasma facing components for the SST-1 tokamak

International Nuclear Information System (INIS)

Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

2000-01-01

Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Plasma core electron density and temperature measurements using CVI line emissions in TCABR Tokamak

Energy Technology Data Exchange (ETDEWEB)

Nascimento, F. do, E-mail: fellypen@ifi.unicamp.br [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Centro de Componentes Semicondutores; Machida, M. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Instituto de Fisica Gleb Wataghin; Severo, J.H.F.; Sanada, E.; Ronchi, G. [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica

2015-08-15

In this work, we present results of electron temperature (T{sub e} ) and density (n {sub e} ) measurements obtained in Tokamak Chauffage Alfven Bresilien (TCABR) tokamak using visible spectroscopy from CVI line emissions which occurs mainly near the center of the plasma column. The presented method is based on a well-known relationship between the particle flux (Γ {sub ion}) and the photon flux (ø {sub ion}) emitted by an ion species combined with ionizations per photon atomic data provided by the atomic data and analysis structure (ADAS) database. In the experiment, we measured the photon fluxes of three different CVI spectral line emissions, 4685.2, 5290.5, and 6200.6 Å (one line per shot). Using this method it was possible to find out the temporal evolution of T{sub e} and n{sub e} in the plasma. The results achieved are in good agreement with T{sub e} and n{sub e} measurements made using other diagnostic tools. (author)

A quasi-linear gyrokinetic transport model for tokamak plasmas

International Nuclear Information System (INIS)

Casati, A.

2009-10-01

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

Design considerations of modular pump limiters for large tokamaks

International Nuclear Information System (INIS)

Uckan, T.; Klepper, C.C.; Mioduszewski, P.K.

1987-11-01

Long-pulse (>10-s) and high-power (>10-MW) operation of large tokamaks requires multiple limiter modules for particle and heat removal, and the power load must be distributed among a number of modules. Because each added module changes the performance of all the others, a set of design criteria must be defined for the overall limiter system. The relationship between individual modules must also be considered from the standpoint of flux coverage and shadowing effects. This paper addresses these issues and provides design guidelines. Parameters of the individual modules are then determined from the system requirements for particle and power removal. Long-pulse operation of large tokamaks requires that the limiter modules be equipped with active cooling. At the leading edge of a module, the cooling channel determines the thickness of the limiter blade (or head). A model has been developed for estimating the system exhaust efficiency in terms of the parameters of the leading edge (i.e., its thickness and the design heat flux) in terms of given device parameters and the power load that must be removed. The impact on module design of state-of-the-art engineering technology for high heat removal is discussed. The choice of locations for the modules is also investigated, and the effects of shadowing between modules on particle and power removal are examined. The results are applied to the Tore Supra tokamak. Conceptual design parameters of the modular pump limiter system are given. 10 refs., 5 figs

Development and validation of a tokamak skin effect transformer model

International Nuclear Information System (INIS)

Romero, J.A.; Moret, J.-M.; Coda, S.; Felici, F.; Garrido, I.

2012-01-01

A lumped parameter, state space model for a tokamak transformer including the slow flux penetration in the plasma (skin effect transformer model) is presented. The model does not require detailed or explicit information about plasma profiles or geometry. Instead, this information is lumped in system variables, parameters and inputs. The model has an exact mathematical structure built from energy and flux conservation theorems, predicting the evolution and non-linear interaction of plasma current and internal inductance as functions of the primary coil currents, plasma resistance, non-inductive current drive and the loop voltage at a specific location inside the plasma (equilibrium loop voltage). Loop voltage profile in the plasma is substituted by a three-point discretization, and ordinary differential equations are used to predict the equilibrium loop voltage as a function of the boundary and resistive loop voltages. This provides a model for equilibrium loop voltage evolution, which is reminiscent of the skin effect. The order and parameters of this differential equation are determined empirically using system identification techniques. Fast plasma current modulation experiments with random binary signals have been conducted in the TCV tokamak to generate the required data for the analysis. Plasma current was modulated under ohmic conditions between 200 and 300 kA with 30 ms rise time, several times faster than its time constant L/R ≈ 200 ms. A second-order linear differential equation for equilibrium loop voltage is sufficient to describe the plasma current and internal inductance modulation with 70% and 38% fit parameters, respectively. The model explains the most salient features of the plasma current transients, such as the inverse correlation between plasma current ramp rates and internal inductance changes, without requiring detailed or explicit information about resistivity profiles. This proves that a lumped parameter modelling approach can be used to

Liquid tin limiter for FTU tokamak

Energy Technology Data Exchange (ETDEWEB)

Vertkov, A., E-mail: avertkov@yandex.ru [JSC “Red Star”, Moscow (Russian Federation); Lyublinski, I. [JSC “Red Star”, Moscow (Russian Federation); NRNU MEPhI, Moscow (Russian Federation); Zharkov, M. [JSC “Red Star”, Moscow (Russian Federation); Mazzitelli, G.; Apicella, M.L.; Iafrati, M. [Associazione EURATOM-ENEA sulla Fusione, C. R. Frascati, Frascati, Rome, Italy, (Italy)

2017-04-15

Highlights: • First steady state operating liquid tin limiter TLL is under study on FTU tokamak. • The cooling system with water spray coolant for TLL has been developed and tested. • High corrosion resistance of W and Mo in molten Sn confirmed up to 1000 °C. • Wetting process with Sn has been developed for Mo and W. - Abstract: The liquid Sn in a matrix of Capillary Porous System (CPS) has a high potential as plasma facing material in steady state operating fusion reactor owing to its physicochemical properties. However, up to now it has no experimental confirmation in tokamak conditions. First steady state operating limiter based on the CPS with liquid Sn installed on FTU tokamak and its experimental study is in progress. Several aspects of the design, structural materials and operation parameters of limiter based on tungsten CPS with liquid Sn are considered. Results of investigation of corrosion resistance of Mo and W in Sn and their wetting process are presented. The heat removal for limiter steady state operation is provided by evaporation of flowing gaswater spray. The effectiveness of such heat removal system is confirmed in modelling tests with power flux up to 5 MW/m2.

TEMPEST simulations of the plasma transport in a single-null tokamak geometry

International Nuclear Information System (INIS)

Xu, X.Q.; Cohen, R.H.; Rognlien, T.D.; Bodi, K.; Krasheninnikov, S.

2010-01-01

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.

Equilibrium reconstruction in the TCA/Br tokamak

International Nuclear Information System (INIS)

Sa, Wanderley Pires de

1996-01-01

The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author)

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

Stochastic layer scaling in the two-wire model for divertor tokamaks

Science.gov (United States)

Ali, Halima; Punjabi, Alkesh; Boozer, Allen

2009-06-01

The question of magnetic field structure in the vicinity of the separatrix in divertor tokamaks is studied. The authors have investigated this problem earlier in a series of papers, using various mathematical techniques. In the present paper, the two-wire model (TWM) [Reiman, A. 1996 Phys. Plasmas 3, 906] is considered. It is noted that, in the TWM, it is useful to consider an extra equation expressing magnetic flux conservation. This equation does not add any more information to the TWM, since the equation is derived from the TWM. This equation is useful for controlling the step size in the numerical integration of the TWM equations. The TWM with the extra equation is called the flux-preserving TWM. Nevertheless, the technique is apparently still plagued by numerical inaccuracies when the perturbation level is low, resulting in an incorrect scaling of the stochastic layer width. The stochastic broadening of the separatrix in the flux-preserving TWM is compared with that in the low mn (poloidal mode number m and toroidal mode number n) map (LMN) [Ali, H., Punjabi, A., Boozer, A. and Evans, T. 2004 Phys. Plasmas 11, 1908]. The flux-preserving TWM and LMN both give Boozer-Rechester 0.5 power scaling of the stochastic layer width with the amplitude of magnetic perturbation when the perturbation is sufficiently large [Boozer, A. and Rechester, A. 1978, Phys. Fluids 21, 682]. The flux-preserving TWM gives a larger stochastic layer width when the perturbation is low, while the LMN gives correct scaling in the low perturbation region. Area-preserving maps such as the LMN respect the Hamiltonian structure of field line trajectories, and have the added advantage of computational efficiency. Also, for a $1\\frac12$ degree of freedom Hamiltonian system such as field lines, maps do not give Arnold diffusion.

Heat flux expressions that satisfy the conservation laws in atomistic system involving multibody potentials

Energy Technology Data Exchange (ETDEWEB)

Fu, Yao, E-mail: Yao.Fu@colorado.edu; Song, Jeong-Hoon, E-mail: JH.Song@colorado.edu

2015-08-01

Heat flux expressions are derived for multibody potential systems by extending the original Hardy's methodology and modifying Admal & Tadmor's formulas. The continuum thermomechanical quantities obtained from these two approaches are easy to compute from molecular dynamics (MD) results, and have been tested for a constant heat flux model in two distinctive systems: crystalline iron and polyethylene (PE) polymer. The convergence criteria and affecting parameters, i.e. spatial and temporal window size, and specific forms of localization function are found to be different between the two systems. The conservation of mass, momentum, and energy are discussed and validated within this atomistic–continuum bridging.

Heat flux expressions that satisfy the conservation laws in atomistic system involving multibody potentials

International Nuclear Information System (INIS)

Fu, Yao; Song, Jeong-Hoon

2015-01-01

Heat flux expressions are derived for multibody potential systems by extending the original Hardy's methodology and modifying Admal & Tadmor's formulas. The continuum thermomechanical quantities obtained from these two approaches are easy to compute from molecular dynamics (MD) results, and have been tested for a constant heat flux model in two distinctive systems: crystalline iron and polyethylene (PE) polymer. The convergence criteria and affecting parameters, i.e. spatial and temporal window size, and specific forms of localization function are found to be different between the two systems. The conservation of mass, momentum, and energy are discussed and validated within this atomistic–continuum bridging

Development of a locally mass flux conservative computer code for calculating 3-D viscous flow in turbomachines

Science.gov (United States)

Walitt, L.

1982-01-01

The VANS successive approximation numerical method was extended to the computation of three dimensional, viscous, transonic flows in turbomachines. A cross-sectional computer code, which conserves mass flux at each point of the cross-sectional surface of computation was developed. In the VANS numerical method, the cross-sectional computation follows a blade-to-blade calculation. Numerical calculations were made for an axial annular turbine cascade and a transonic, centrifugal impeller with splitter vanes. The subsonic turbine cascade computation was generated in blade-to-blade surface to evaluate the accuracy of the blade-to-blade mode of marching. Calculated blade pressures at the hub, mid, and tip radii of the cascade agreed with corresponding measurements. The transonic impeller computation was conducted to test the newly developed locally mass flux conservative cross-sectional computer code. Both blade-to-blade and cross sectional modes of calculation were implemented for this problem. A triplet point shock structure was computed in the inducer region of the impeller. In addition, time-averaged shroud static pressures generally agreed with measured shroud pressures. It is concluded that the blade-to-blade computation produces a useful engineering flow field in regions of subsonic relative flow; and cross-sectional computation, with a locally mass flux conservative continuity equation, is required to compute the shock waves in regions of supersonic relative flow.

Determination of 3D Equilibria from Flux Surface Knowledge Only

International Nuclear Information System (INIS)

Mynick, H.E.; Pomphrey, N.

2001-01-01

We show that the method of Christiansen and Taylor, from which complete tokamak equilibria can be determined given only knowledge of the shape of the flux surfaces, can be extended to 3-dimensional equilibria, such as those of stellarators. As for the tokamak case, the given geometric knowledge has a high degree of redundancy, so that the full equilibrium can be obtained using only a small portion of that information

Tokamak plasma shape identification based on the boundary integral equations

International Nuclear Information System (INIS)

Kurihara, Kenichi; Kimura, Toyoaki

1992-05-01

A necessary condition for tokamak plasma shape identification is discussed and a new identification method is proposed in this article. This method is based on the boundary integral equations governing a vacuum region around a plasma with only the measurement of either magnetic fluxes or magnetic flux intensities. It can identify various plasmas with low to high ellipticities with the precision determined by the number of the magnetic sensors. This method is applicable to real-time control and visualization using a 'table-look-up' procedure. (author)

Control of plasma poloidal shape and position in the DIII-D tokamak

International Nuclear Information System (INIS)

Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

1997-11-01

Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

Divertor layer during the thermal phase of the disruption in a tokamak

International Nuclear Information System (INIS)

Konkashbaev, I.K.

1993-01-01

Physical processes in the tokamak divertor with a poloidal field during plasma disruption are considered. It is shown that the processes differ qualitatively from the processes in a stationary mode. During the disruption the energy flux is increased by 10 4 times

Kinetic and transport theory near the tokamak edge

International Nuclear Information System (INIS)

Hazeltine, R.D.; Catto, P.J.

1995-12-01

Conventional transport orderings employed in the core of a tokamak plasma allow large divergence-free flows in flux surfaces, but only weak radial flows. However, alternate orderings are required in the edge region where radial diffusion must balance the rapid loss due to free-streaming to divertor plates or limiters. Kinetic equations commonly used to study the plasma core do not allow such a balance and are, therefore, inapplicable in the plasma edge. Similarly, core transport formulae cannot be extended to the edge region without major, qualitative alteration. Here the authors address the necessary changes. By deriving and solving a novel kinetic equation, they construct distinctive collisional transport laws for the plasma edge. They find that their edge ordering naturally retains the radial diffusion and parallel flow of particles, momentum and heat to lowest order in the conservation equations. To higher order they find a surprising form for parallel transport in the scrape-off layer, in which the parallel flow of particles and heat are driven by a combination of the conventional gradients, viscosity, and new terms involving radial derivatives. The new terms are not relatively small, and could affect understanding of limiter and divertor operation

Theoretical issues in tokamak confinement: (i) Internal/edge transport barriers and (ii) runaway avalanche confinement

International Nuclear Information System (INIS)

Connor, J.W.; Helander, P.; Thyagaraja, A.; Andersson, F.; Fueloep, T.; Eriksson, L.-G.; Romanelli, M.

2001-01-01

This paper summarises a number of distinct, but related, pieces of work on key confinement issues for tokamaks, in particular the formation of internal and edge transport barriers, both within turbulent and neoclassical models, and radial diffusion of avalanching runaway electrons. First-principle simulations of tokamak turbulence and transport using the two-fluid, electromagnetic, global code CUTIE are described. The code has demonstrated the spontaneous formation of internal transport barriers near mode rational surfaces, in qualitative agreement with observations on JET and RTP. The theory of neoclassical transport in an impure, toroidal plasma has been extended to allow for steeper pressure and temperature gradients than are usually considered, and is then found to become nonlinear under conditions typical of the tokamak edge. For instance, the particle flux is found to be a nonmonotonic function of the gradients, thus allowing for a bifurcation in the ion particle flux. Finally, it is shown that radial diffusion caused by magnetic fluctuations can effectively suppress avalanches of runaway electrons if the fluctuation amplitude exceeds δB/B∼10 -3 . (author)

A model for plasma discharges simulation in Tokamak devices

International Nuclear Information System (INIS)

Fonseca, Antonio M.M.; Silva, Ruy P. da; Galvao, Ricardo M.O.; Kusnetzov, Yuri; Nascimento, I.C.; Cuevas, Nelson

2001-01-01

In this work, a 'zero-dimensional' model for simulation of discharges in Tokamak machine is presented. The model allows the calculation of the time profiles of important parameters of the discharge. The model was applied to the TCABR Tokamak to study the influence of parameters and physical processes during the discharges. Basically it is constituted of five differential equations: two related to the primary and secondary circuits of the ohmic heating transformer and the other three conservation equations of energy, charge and neutral particles. From the physical model, a computer program has been built with the objective of obtaining the time profiles of plasma current, the current in the primary of the ohmic heating transformer, the electronic temperature, the electronic density and the neutral particle density. It was also possible, with the model, to simulate the effects of gas puffing during the shot. The results of the simulation were compared with the experimental results obtained in the TCABR Tokamak, using hydrogen gas

L-mode SOL width scaling in the MAST spherical tokamak

International Nuclear Information System (INIS)

Ahn, J-W; Counsell, G F; Kirk, A

2006-01-01

A new data-set of outboard mid-plane scrape-off layer (SOL) heat flux widths, Δ h , has been constructed for L-mode plasmas in the MAST spherical tokamak (ST). The scaling with key plasma parameters such as density, toroidal magnetic field, parallel connection length in the SOL and surface heat flux at the separatrix is investigated. An empirical scaling is developed for the Δ h data-set, which exhibits a strong positive dependence on both the connection length (or edge safety factor) and density and weak or moderate inverse dependences on the surface heat flux and magnetic field, respectively. The empirical scaling is compared with earlier results for a range of tokamaks with conventional geometry, which show weaker dependence on the density and edge safety factor. Importantly, however, the weak negative dependence on the surface heat flux (and thus heating power) is common in both conventional and ST geometries. The experimental data are also used to test a number of dimensionally correct Δ h scalings developed from theoretical models for perpendicular transport in the SOL coupled with classical transport parallel to the magnetic field. A scaling based on perpendicular transport driven by resistive MHD interchange provides the best fit, although several models are close. A subset of the better fitting theoretical scalings are used to extrapolate for Δ h in one design for a future burning ST machine and finally to predict the peak heat loading on the outboard divertor target plate

The effects of plasma deformability on the feedback stabilization of axisymmetric modes in tokamak plasmas

International Nuclear Information System (INIS)

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

1991-09-01

The effects of plasma deformability on the feedback stabilization of axisymmetric modes of tokamak plasmas are studied. It is seen that plasmas with strongly shaped cross sections have unstable motion different from a rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the eigenfunction in a way that reduces the stabilizing eddy currents in these conductors. Passive feedback results using several equilibria of varying shape are presented. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops which are used to determine plasma vertical position for the active feedback system. The variations of these non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations. Active feedback results are presented for the PBX-M tokamak configuration. (author) 19 figs., 2 tabs., 30 refs

Design and experimental results of feedback control of Ohmic-heating transformer magnetic flux by LHCD power in HT-7 Tokamak

International Nuclear Information System (INIS)

Yiyun Huang

2006-01-01

In order to make a research on long pulse or even steady state operation with non-inductive drive in plasma discharge, a new feedback control scheme instead of the previous one has been designed and operated in HT-7 [HT-7 team presented by J. Li, et al., Plasma Phys. Control. Fusion 42 (2) (2000) 135-146] Tokamak experiment, 2004. Consumption of iron-core transformer magnetic flux (MFT) is feedback controlled for the first time by power of lower hybrid current drive (LHCD) P LH , when the Ohmic-heating circuit current can maintain the plasma current I P constant with another feedback control loop, which make MFT evolve at alternating-change state to avoid flux saturation. Plasma current I P can be maintained steadily up to 120s in this operation mode at reduced plasma parameters (I P ∼50-100KA, average density n-bar e =0.4-0.5x10 19 m -3 , P LH =100-200KW). Design and experimental results are presented in the paper, which including control model analysis, configurations of control system and MFT feedback control experiments in HT-7. The high voltage power supply (HVPS) of LHCD is the main controller that regulates the LHCD power into the plasma to control the MFT

Energy selecting action of limiters on particle fluxes penetrating into the SOL-plasma

International Nuclear Information System (INIS)

Hildebrandt, D.

1986-01-01

A single model of the penetration of particle effluxes from the core plasma into the SOL-plasma of tokamaks is proposed. The assumptions made are free streaming of particles parallel to the magnetic field and anomalous particle transport perpendicular to the toroidal field with a constant radial velocity. The model has been proved for measured particle fluxes of Li which was injected into the core plasma of the tokamak T-10. The dependence of the Li-particle flux on the minor radius as well as toroidal asymmetries in the SOL-plasma can be explained by the results of the calculations. (author)

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

International Nuclear Information System (INIS)

Kiviniemi, T.

2001-01-01

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

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.

Scrape-off layer width of parallel heat flux on tokamak COMPASS

Czech Academy of Sciences Publication Activity Database

Loureiro, J.; Silva, C.; Horáček, Jan; Adámek, Jiří; Stöckel, Jan

2014-01-01

Roč. 1, č. 3 (2014), s. 121 ISSN 2336-2626. [SPPT 2014 - 26th Symposium on Plasma Physics and Technology/26./. Prague, 16.06.2014-19.06.2014] R&D Projects: GA ČR(CZ) GAP205/12/2327 Institutional support: RVO:61389021 Keywords : tokamak * edge turbulent transport * Scrape-Off layer * Langmuir probe * Ball- pen probe Subject RIV: BL - Plasma and Gas Discharge Physics http://fyzika.feld.cvut.cz/misc/ppt/articles/2014/loureiro.pdf

Tokamak Systems Code

International Nuclear Information System (INIS)

Reid, R.L.; Barrett, R.J.; Brown, T.G.

1985-03-01

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

Asymmetry of edge plasma turbulence in biasing experiments on tokamak TF-2

International Nuclear Information System (INIS)

Budaev, V.P.

1994-01-01

It was observed in tokamaks the suppression of edge turbulence causes by setting a radial electric field at the plasma edge. The poloidal plasma rotation governed by this electric field is likely to result in changes in edge convention and poloidal asymmetry, however there is no experimental evidence about that of the experimental database concerning the biasing and conditions of edge plasma electrostatic turbulence excitation is not still complete. Also a relation between macroscopic convection and small-scale electrostatic turbulence have not yet revealed both in biasing and non biasing plasmas. In this paper results from biasing experiments carried on on ohmically heated tokamak TF-2 are presented. Changes in both equilibrium and fluctuated edge plasma parameters also convection and turbulence driven particle flux were demonstrated in probe measurements with biasing of electrode immersed within Last Closed Flux Surface (LCFS). Poloidal edge plasma structure and charge in asymmetry have demonstrated in the biasing experiments. (author). 6 refs, 4 figs

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)

Neoclassical alpha-particle losses in tokamaks allowing for large orbit widths

International Nuclear Information System (INIS)

Cox, M.; O'Brien, M.R.; Zaitsev, F.S.

1994-01-01

Alpha-particle physics is of particular importance now that research into controlled fusion has reached thermonuclear parameters and D-T fuel has been used in JET and TFTR. Here we address the important topic of α-particle transport: if transport is too low helium ash accumulates quenching the burn; if it is too high heating of the plasma by fast α-particles is insufficient to maintain the burn. We give results from simulations of α-particle distributions (f α ) which self-consistently treat α-particle birth, collisional slowing down and neoclassical radial transport. The (steady-state) f α is calculated by the FPP code as a function of speed (v), pitch-angle (θ) and flux surface radius (r). This code is based on a 3D Fokker-Planck theory of 'banana regime' neoclassical effects in tokamaks which can treat large deviations of fast ion orbits from flux surfaces and non-Maxwellian distributions. The code reproduces standard neoclassical results for Maxwellian distributions in the large aspect ratio (ε) and small orbit width (Δ) limits (e.g. radial fluxes, conductivities and bootstrap currents), but can also be used for small ε and large Δ which are difficult to treat analytically. The code is particularly useful for α-particle studies as (a) the experimental evidence is that fast ion transport is usually consistent with neoclassical theory, unlike electron or thermal ion transport, and (b) trapped fast ion orbits can deviate greatly from flux surfaces. An alternative to this Fokker-Planck treatment is Monte Carlo modelling. However, representation of the detailed structure of f α (θ,v,r) would require very large number of particles, and hence be very slow. Calculations have been made for parameters typical of TFTR, JET, SSTR (an 'advanced tokamak' reactor) and STR (a tight aspect ratio or 'spherical' tokamak reactor, though only the JET results are discussed in detail. (author) 4 refs., 4 figs

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-08-15

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

Quasilinear kinetic modeling of RMP penetration into a tokamak plasma

International Nuclear Information System (INIS)

Heyn, M.F.; Kernbichler, W.; Leitner, P.; Ivanov, I.B.; Kasilov, S.V.

2013-01-01

The linear as well as the quasilinear problem of RMP penetration in tokamaks is solved consistently with a particle and energy conserving collision operator. The new collision operator ensures the Onsager symmetry of the quasilinear transport coefficient matrix and avoids artifacts such as fake heat convection connected with simplified collision models.

The steady-state tokamak program

International Nuclear Information System (INIS)

Politzer, D.A.; Nevins, W.M.

1992-01-01

This paper reports on a steady-state tokamak experiment (STE) needed to develop the technology and physics data base required for construction of a steady-state fusion power demonstration reactor in the early 21st century. The STE will provide an integrated facility for the development and demonstration of steady-state and particle handling, low-activation high-heat-flux components and materials, efficient current drive, and continuous plasma performance in steady-state, with reactor-like plasma conditions under severe conditions of heat and particle bombardment of the wall. The STE facility will also be used to develop operation and control scenarios for ITER

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

Plasma edge physics in an actively cooled tokamak

International Nuclear Information System (INIS)

Gunn, J.P.; Adamek, A.; Boucher, C.

2005-01-01

Tore Supra is a large tokamak with a plasma of circular cross section (major radius 2.4 m and minor radius 0.72 m) lying on a toroidal limiter. Tore Supra's main mission is the development of technology to inject up to 25 MW of microwave heating power and extract it continuously for up to 1000 s in steady state without uncontrolled overheating of, or outgassing from, plasma-facing components. The entire first wall of the tokamak is actively cooled by a high pressure water loop and special carbon fiber composite materials have been designed to handle power fluxes up to 10 MW/m 2 . The edge plasma on open magnetic flux surfaces that intersect solid objects plays an important role in the overall behaviour of the plasma. The transport of sputtered impurity ions and the fueling of the core plasma are largely governed by edge plasma density, temperature, and flow profiles. Measurements of these quantities are becoming more reliable and frequent in many tokamaks, and it has become clear that we do not understand them very well. Classical two-dimensional fluid modelling fails to reproduce many aspects of the experimental observations such as the significant thickness of the edge plasma, and the near-sonic flows that occur where none should be expected. It is suspected that plasma turbulence is responsible for these anomalies. In the Tore Supra tokamak, various kinds of Langmuir probes are used to characterize the edge plasma. We will present original measurements that demonstrate the universality of many phenomena that have been observed in X-point divertor tokamaks, especially concerning the ion flows. As in the JET tokamak, surprisingly large values of parallel Mach number are measured midway between the two strike zones, where one would expect to find nearly stagnant plasma if the particle source were poloidally uniform. We will present results of a novel experiment that provides evidence for a poloidally localized particle and energy source on the outboard midplane of

Low Z impurity transport in tokamaks. [Neoclassical transport theory

Energy Technology Data Exchange (ETDEWEB)

Hawryluk, R.J.; Suckewer, S.; Hirshman, S.P.

1978-10-01

Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

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

Thermal loads on tokamak plasma-facing components during normal operation and disruptions

International Nuclear Information System (INIS)

McGrath, R.T.

1990-01-01

Power loadings experienced by tokamak plasma-facing components during normal operation and during off-normal events are discussed. A model for power and particle flow in the tokamak boundary layer is presented and model predictions are compared to infrared measurements of component heating. The inclusion of the full three-dimensional geometry of the components and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armour tile surfaces by only a millimeter can lead to significant localized heating. An application to the design of plasma-facing components for future machines is presented. Finally, thermal loads expected during tokamak disruptions are discussed. The primary problems are surface melting and vaporization due to localized intense heating during the disruption thermal quench and volumetric heating of the component armour and structure due to localised impact of runaway electrons. (author)

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.

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

Heat flux driven ion turbulence

International Nuclear Information System (INIS)

Garbet, X.

1998-01-01

This work is an analysis of an ion turbulence in a tokamak in the case where the thermal flux is fixed and the temperature profile is allowed to fluctuate. The system exhibits some features of Self-Organized Critical systems. In particular, avalanches are observed. Also the frequency spectrum of the thermal flux exhibits a structure similar to the one of a sand pile automaton, including a 1/f behavior. However, the time average temperature profile is found to be supercritical, i.e. the temperature gradient stays above the critical value. Moreover, the heat diffusivity is lower for a turbulence calculated at fixed flux than a fixed temperature gradient, with the same time average temperature. This behavior is attributed to a stabilizing effect of avalanches. (author)

Problems with the concept of plasma equilibrium in tokamaks

International Nuclear Information System (INIS)

Carreras, B.A.

1992-01-01

The equilibrium condition for a magnetically confined plasma in normally formulated in terms of macroscopic equations. In these equations, the plasma pressure is assumed to be a function of the magnetic flux with continuous derivatives. However, in three- dimensional systems this is not necessarily the case. Here, we look at the case of an intrinsically three-dimensional realistic tokamak, and we discuss the possible interconnection between the equilibrium and anomalous transport

Studies of feedback stabilization of axisymmetric modes in deformable tokamak plasmas

International Nuclear Information System (INIS)

Ward, D.J.

1991-01-01

A new linear MHD stability code, NOVA-W, is described and applied to the study of the feedback stabilization of the axisymmetric mode in deformable tokamak plasma. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The code has been tested for the case of passive stabilization against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. The NOVA-W code is used to examine the effects of plasma deformability on feedback stabilization. It is seen that plasmas with shaped cross sections have unstable motion different from a rigid shift. Plasma equilibria with large triangularity show particularly significant deviations from a uniform rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the motion in a way that reduces the stabilizing effects of these conductors. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops. These non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations

Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

Energy Technology Data Exchange (ETDEWEB)

Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

2016-12-15

Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

Simulations of Turbulence in Tokamak Edge and Effects of Self-Consistent Zonal Flows

Science.gov (United States)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge. This extends previous work to include self-consistent zonal flows and their effects. The previous work addressed simulation of L-mode tokamak edge turbulence using the turbulence code BOUT that solves Braginskii-based plasma fluid equations in tokamak edge domain. The calculations use realistic single-null geometry and plasma parameters of the DIII-D tokamak and produce fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Thermographic analysis of plasma facing components covered by carbon surface layer in tokamaks

International Nuclear Information System (INIS)

Gardarein, Jean-Laurent

2007-01-01

Tokamaks are reactors based on the thermonuclear fusion energy with magnetic confinement of the plasma. In theses machines, several MW are coupled to the plasma for about 10 s. A large part of this power is directed towards plasma facing components (PFC). For better understanding and control the heat flux transfer from the plasma to the surrounding wall, it is very important to measure the surface temperature of the PFC and to estimate the imposed heat flux. In most of tokamaks using carbon PFC, the eroded carbon is circulating in the plasma and redeposited elsewhere. During the plasma operations, this leads at some locations to the formation of thin or thick carbon layers usually poorly attached to the PFC. These surface layers with unknown thermal properties complicate the calculation of the heat flux from IR surface temperature measurements. To solve this problem, we develop first, inverse method to estimate the heat flux using thermocouple (not sensitive to the carbon surface layers) temperature measurements. Then, we propose a front face pulsed photothermal method allowing an estimation of layers thermal diffusivity, conductivity, effusivity and the thermal contact resistance between the layer and the tile. The principle is to study with an infrared sensor, the cooling of the layer surface after heating by a short laser pulse, this cooling depending on the thermal properties of the successive layers. (author) [fr

Viscous damping of toroidal angular momentum in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Stacey, W. M. [Georgia Tech Fusion Research Center, Atlanta, Georgia 30332 (United States)

2014-09-15

The Braginskii viscous stress tensor formalism was generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry in order to provide a representation for the viscous damping of toroidal rotation in tokamaks arising from various “neoclassical toroidal viscosity” mechanisms. In the process, it was verified that the parallel viscosity contribution to damping toroidal angular momentum still vanishes even in the presence of toroidal asymmetries, unless there are 3D radial magnetic fields.

Reclosing of field lines and disruptive instability in tokamaks

International Nuclear Information System (INIS)

Kadomtsev, B.B.

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

Low-temperature operating regime of the tokamak evacuating limiter

International Nuclear Information System (INIS)

Tokar', M.Z.

1987-01-01

The conditions for realizing the regime of strong recycling of a cold dense plasma of an evacuating limiter were determined based on a previously proposed model for describing the limiter layer of a tokamak. The scaling for the dependence of the gas pressure in the evacuation system on the average plasma density in the limiter layer was found, and agreed quantitatively with the results of measurements on the Alcator and ISX-B tokamaks. For the tokamak reactor of the INTOR scale the calculations show that the low-temperature operating regime of the evacuating limiter can be realized with a quite low pumping rate. It has the advantages of reduced erosion of the limiter and small fluxes of impurities into the working volume of the reactor. In addition, the relative concentration of the helium ash in the limiter layer does not exceed 2-3%, but the density of the main plasma is comparable to the proposed average density in the reactor. The concept of a stochastic limiter is of interest for lowering the plasma density in the limiter layer and lowering the thermal loads on the limiter

Reclosing of field lines and disruptive instability in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Kadomtsev, B. B.

1976-07-01

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

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

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

Electromagnetic forces on a metallic Tokamak vacuum vessel following a disruptive instability

International Nuclear Information System (INIS)

Eckhartt, D.

1979-04-01

During a 'hard' disruptive instability of a Tokamak plasma the current-carrying plasma is lost within a very short time, typically few milliseconds. If the plasma is contained in a metallic vacuum vessel, electric currents are set up in the vessel following the disappearance of the plasma current. These vessel currents together with the magnetic fields intersecting the vessel generate electromagnetic forces which appear as mechanical loads on the vessel. In the following note it is assumed that the vacuum vessel is surrounded by an 'outer equivalent' or 'flux-conserving' shell having a characteristic time of magnetic field penetration which is long compared to the time of existence of the vessel currents. This property defines the distribution of vessel current densities (and hence the load distribution) without referring to the exact mechanism or time sequence of events by which the plasma current is lost. Numerical examples of the electromagnetic force distribution from this model refer to parameters of the JET-device with the simplifying assumption of circular cross-sections for plasma current, vacuum vessel, and outer equivalent shell. (orig.)

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

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

Nonrigid, Linear Plasma Response Model Based on Perturbed Equilibria for Axisymmetric Tokamak Control Design

International Nuclear Information System (INIS)

Welander, A.S.; Deranian, R.D.; Humphreys, D.A.; Leuer, J.A.; Walker, M.L.

2005-01-01

Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII-D, the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone.In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. The linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution

Integrated modeling of temperature profiles in L-mode tokamak discharges

Energy Technology Data Exchange (ETDEWEB)

Rafiq, T.; Kritz, A. H.; Tangri, V. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Voitsekhovitch, I. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Budny, R. V. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2014-12-15

Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows

Science.gov (United States)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Radial diffusion of a minority species in a tokamak due to ICRH

International Nuclear Information System (INIS)

Vacca, L.

1993-01-01

The author studies the transport of a minority species in a scenario where minority ions in a tokamak are heated by fast Alfven waves having a resonance layer in a tokamak. He does not assume the minority distribution function to be a Maxwellian at leading order, as transport theory generally assumes, but adopts a more realistic model where the strong anisotropy of the distribution function is accounted for. This anisotropy has been observed in experiments and is predicted by numerical calculations based on Fokker-Planck equation with quasilinear diffusion. By adopting a different ordering from that used in previous work on transport due to waves and taking moments of the kinetic equation which includes the rf driving term, he calculates the fluxes of the resonant minority species accounting for collisions of minority with bulk electrons and ion species. Finally he makes comparison of fluxes of rf-heated minority with standard neoclassical predictions (no rf source present) showing the enhancement in transport introduced by the presence of both a strongly anisotropic distribution function and an rf source

Peeling-off of the external kink modes at tokamak plasma edge

International Nuclear Information System (INIS)

Zheng, L. J.; Furukawa, M.

2014-01-01

It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture

Peeling-off of the external kink modes at tokamak plasma edge

Energy Technology Data Exchange (ETDEWEB)

Zheng, L. J. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States); Furukawa, M. [Graduate School of Engineering, Tottori University, Tottori 680-8552 (Japan)

2014-08-15

It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture.

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.)

Hot spots on the neutralizer plates of a tokamak

International Nuclear Information System (INIS)

Krasheninnikov, S.I.

1991-01-01

The formation of hot spots on the neutralizing surfaces of tokamaks may be one of the reasons for the entry of large impurity fluxes into the plasmas of TFTR and JET (the so-called carbon catastrophe or carbon bloom) with high auxiliary heating powers. At this time it is unclear whether these hot spots are caused just by nonuniformities on the neutralizer surface or whether their appearance is the result of some more general behavior, with the surface nonuniformities only showing up as seed perturbations. In this paper it is shown that hot spots can also develop on smooth surfaces of carbon neutralizer plates as a result of the contraction of a heat flux incident on the plates

Joint research using small tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

Well-balanced schemes for the Euler equations with gravitation: Conservative formulation using global fluxes

Science.gov (United States)

Chertock, Alina; Cui, Shumo; Kurganov, Alexander; Özcan, Şeyma Nur; Tadmor, Eitan

2018-04-01

We develop a second-order well-balanced central-upwind scheme for the compressible Euler equations with gravitational source term. Here, we advocate a new paradigm based on a purely conservative reformulation of the equations using global fluxes. The proposed scheme is capable of exactly preserving steady-state solutions expressed in terms of a nonlocal equilibrium variable. A crucial step in the construction of the second-order scheme is a well-balanced piecewise linear reconstruction of equilibrium variables combined with a well-balanced central-upwind evolution in time, which is adapted to reduce the amount of numerical viscosity when the flow is at (near) steady-state regime. We show the performance of our newly developed central-upwind scheme and demonstrate importance of perfect balance between the fluxes and gravitational forces in a series of one- and two-dimensional examples.

Modification of boundary fluctuations by LHCD in the HT-7 tokamak

International Nuclear Information System (INIS)

Song Mei; Wan Baonian; Xu Guosheng; Ling Bili

2003-01-01

Measurements of boundary fluctuations and fluctuation driven electron fluxes have been performed in ohmic and lower hybrid current drive enhanced confinement plasma using a graphite Langmuir probe array on HT-7 tokamak. The fluctuations are significantly suppressed and the turbulent fluxes are remarkably depressed in the enhanced plasma. We characterized the statistical properties of fluctuations and the particle flux and found a non-Gaussian character in the whole scrape-off layer with minimum deviations from Gaussian in the proximity of the velocity shear layer in ohmic plasma. In the enhanced plasma the deviations in the boundary region are all reduces obviously. The fluctuations and induced electron fluxes show sporadic bursts asymmetric in time and the asymmetry is remarkably weakened in the lower hybrid current driving (LHCD) phase. The results suggest a coupling between the statistical behaviour of fluctuations and the turbulent flow

Beat-wave excitation and current driven in tokamak plasma. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Mohamed, B F [Plasma physics Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

Wave heating current drive in tokamaks is a growing subject in the plasma physics literature. For current drive in tokamaks by electromagnetic waves, different methods have been proposed recently. One of the promising schemes for current drive remains the beat wave scheme. This technique employs two CO- or counterpropagating monochromatic laser beams (or microwaves) whose frequency difference matches the plasma frequency, while the wave number difference (or sum, in the case of counterpropagating) determine the wave number of the resulting plasma beat wave. In this work, the basic analysis of a beat wave current drive scheme in which collinear waves are used is discussed. by assuming a Gaussian profile for the amplitude of these pump waves, the amplitudes of the longitudinal and radial fields of the beat wave due to the nonlinear wave interactions have been calculated. Besides, the transfer of momentum flux that accompanies the transfer of wave action in beat-wave scattering will be used to drive the toroidal radial currents in tokamaks. self-generated magnetic fields due to those currents were also calculated. 1 fig.

Material impacts and heat flux characterization of an electrothermal plasma source with an applied magnetic field

Science.gov (United States)

Gebhart, T. E.; Martinez-Rodriguez, R. A.; Baylor, L. R.; Rapp, J.; Winfrey, A. L.

2017-08-01

To produce a realistic tokamak-like plasma environment in linear plasma device, a transient source is needed to deliver heat and particle fluxes similar to those seen in an edge localized mode (ELM). ELMs in future large tokamaks will deliver heat fluxes of ˜1 GW/m2 to the divertor plasma facing components at a few Hz. An electrothermal plasma source can deliver heat fluxes of this magnitude. These sources operate in an ablative arc regime which is driven by a DC capacitive discharge. An electrothermal source was configured with two pulse lengths and tested under a solenoidal magnetic field to determine the resulting impact on liner ablation, plasma parameters, and delivered heat flux. The arc travels through and ablates a boron nitride liner and strikes a tungsten plate. The tungsten target plate is analyzed for surface damage using a scanning electron microscope.

Joint research using small tokamaks

International Nuclear Information System (INIS)

Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

2005-01-01

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

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

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

1997-01-01

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

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

Sliding Mode Control of a Tokamak Transformer

Energy Technology Data Exchange (ETDEWEB)

Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.

2012-06-08

A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.

Optimization study of normal conductor tokamak for commercial neutron source

Science.gov (United States)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Formation of compact toroidal plasmas by magnetized coaxial plasma gun injection into an oblate flux conserver

International Nuclear Information System (INIS)

Turner, W.C.; Goldenbaum, G.C.; Granneman, E.H.A.; Hartman, C.W.; Prono, D.S.; Taska, J.; Smith, A.C. Jr.

1980-01-01

Initial results are reported on the formation of compact toroidal plasmas in an oblate shaped metallic flux conserver. A schematic of the experimental apparatus is shown. The plasma injector is a coaxial plasma gun with solenoid coils wound on the inner and outer electrodes. The electrode length is 100 cm, the diameter of the inner (outer) electrode is 19.3 cm (32.4 cm). Deuterium gas is puffed into the region between electrodes by eight pulsed valves located on the outer electrode 50 cm from the end of the gun. The gun injects into a cylindrically symmetrical copper shell (wall thickness = 1.6 mm) which acts as a flux conserver for the time scale of experiments reported here. The copper shell consists of a transition cylinder 30 cm long, 34 cm in diameter, a cylindrical oblate pill box 40 cm long, 75 cm in diameter and a downstream cylinder 30 cm long, 30 cm in diameter. The gap between the gun and transition cylinder is 6 cm. An axial array of coils outside the vacuum chamber can be used to establish an initial uniform bias field

Neutral-beam deposition in large, finite-beta noncircular tokamak plasmas

International Nuclear Information System (INIS)

Wieland, R.M.; Houlberg, W.A.

1982-02-01

A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross section and noncircular shifted plasma cross sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is benchmarked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radius a = 120 cm and elongation b/a = 1.6. Deposition profiles are compared for deuterium beam energies of 120 to 150 keV, central plasma densities of 8 x 10 13 - 2 x 10 14 cm -3 , and beam orientation ranging from perpendicular to tangential to the inside wall

Control of particles flux in a tokamak with an events structure

International Nuclear Information System (INIS)

Tsitrone, E.

1995-01-01

Two key problems in the development of a controlled fusion reactor are: -the control of the ashes resulting from the fusion reaction (helium) and of the impurities coming from the wall erosion, which affect the central plasma performances by diluting the fuel and dissipating a part of the produced energy by radiation. - the removal of the heat carried to the walls by charged particles, which is highly concentrated (peak values of several tens of MW per m 2 ). Two types of systems are generally used for the plasma-wall interface: throat limiter and axisymmetric divertor. Neither is an ideal candidate to control simultaneously the heat and particle fluxes. This thesis investigates an alternative configuration, the vented limiter, tested for the first time on the Tore Supra tokamak. The vented limiter principle lies on the recycling neutrals collection by slots, in such a way that local thermal overload is avoided. It is shown experimentally that the surface temperature of the prototype installed in Tore Supra remains uniform. As far as the particle collection is concerned, even though the pressure in the vented limiter is lower than the pressure in the throat limiter by a factor 3 for deuterium and 4 helium, it is sufficient to control the plasma density. Moreover, as with a throat limiter, the pressure exhibits a quadratic evolution with the plasma density. To interpret these results, a model describing the plasma recycling on the limiter and the pumping by the slots has been developed. The model has been validated by a comparison with the experimental data. It was then used to propose an optimized version of the prototype with reshaped slots. This should improve the pumping efficiency by a factor 2, in deuterium as well as in helium, but without removing the discrepancy between both pumping efficiencies. As a consequence, even if the thermal behaviour of the vented limiter is satisfactory, its suitability for a future strongly depends on whether it is possible or

Turbulent regimes in the tokamak scrape-off layer

International Nuclear Information System (INIS)

Mosetto, A.

2014-01-01

The tokamak scrape-off layer (SOL) is the plasma region characterized by open field lines that start and end on the vessel walls. The plasma dynamics in the SOL plays a crucial role in determining the overall performance of a tokamak, since it controls the plasma-wall interactions, being responsible of exhausting the tokamak power, it regulates the overall plasma confinement, and it governs the plasma refueling and the removal of fusion ashes. Scrape-off layer physics is intrinsically non-linear and characterized by phenomena that occur on a wide range of spatio-temporal scales. Free energy sources drive a number of unstable modes that develop into turbulence and lead to transport of particles and heat across the magnetic field lines. Depending on the driving instability, different SOL turbulent regimes can be identified. As the SOL turbulent regimes determine the plasma confinement properties and the SOL width (and, consequently, the power flux on the vessel wall, for example), it is of crucial importance to understand which turbulent regimes are active in the SOL, under which conditions they develop, and which are the main properties of the associated turbulent transport. In the present thesis we define the SOL turbulent regimes, and we provide a framework to identify them, given the operational SOL parameters. Our study is based on the drift-reduced Braginskii equations and it is focused on a limited tokamak SOL configuration. We first describe the main SOL linear instabilities, such as the inertial and resistive branches of the drift waves, the resistive, inertial and ideal branches of the ballooning modes, and the ion temperature gradient mode. Then, we find the SOL turbulent regimes depending on the instability driving turbulent transport, assuming that turbulence saturates when the radial gradient associated to the pressure fluctuations is comparable to the equilibrium one. Our methodology for the turbulent regime identification is supported by the analysis

Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

International Nuclear Information System (INIS)

Abiteboul, J.

2012-10-01

The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are

A flexible software design to determine the plasma boundary in Damavand tokamak

International Nuclear Information System (INIS)

Ghadiri, R.; Esteki, M.H.; Sadeghi, Y.

2014-01-01

A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code “The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)” was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C language. (author)

Flux measurements with a sniffer probe near the wall in ASDEX

International Nuclear Information System (INIS)

Poschenrieder, W.; Venus, G.; Wang, Y.G.; Mueller, E.R.; Bartiromo, R.; Becker, G.; Bosch, H.S.; Brocken, H.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. v.; Glock, E.; Gruber, O.; Haas, G.; Janeschitz, G.; Karger, F.; Kotze, P.B.; Keilhacker, M.; Klueber, O.; Kornherr, M.; Lackner, K.; Lenoci, M.; Lisitano, G.; Mayer, H.M.; McCormick, K.; Meisel, D.; Mertens, V.; Murmann, H.; Niedermeyer, H.; Rapp, H.; Roehr, H.; Ryter, F.; Schneider, F.; Siller, G.; Smeulders, P.; Soeldner, F.; Speth, E.; Steuer, K.H.; Vollmer, O.; Wagner, F.

1985-01-01

For a detailed assessment of particle recycling in a tokamak it is necessary to know quality and quantity of the particle fluxes directed to the elements of the wall. In a divertor machine like ASDEX we have to differentiate between at least four distinct elements: main chamber wall, protective limiters, collector plates, and divertor walls. Relevant data about the divertor region are obtained from pressure and flux measurements. (orig./GG)

Particle and heat transport in Tokamaks

International Nuclear Information System (INIS)

Chatelier, M.

1984-01-01

A limitation to performances of tokamaks is heat transport through magnetic surfaces. Principles of ''classical'' or ''neoclassical'' transport -i.e. transport due to particle and heat fluxes due to Coulomb scattering of charged particle in a magnetic field- are exposed. It is shown that beside this classical effect, ''anomalous'' transport occurs; it is associated to the existence of fluctuating electric or magnetic fields which can appear in the plasma as a result of charge and current perturbations. Tearing modes and drift wave instabilities are taken as typical examples. Experimental features are presented which show that ions behave approximately in a classical way whereas electrons are strongly anomalous [fr

Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma

International Nuclear Information System (INIS)

Selig, G.

2012-12-01

In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)

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

Directory of Open Access Journals (Sweden)

Lin-Liu Y.R.

2012-09-01

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

Flux networks in metabolic graphs

International Nuclear Information System (INIS)

Warren, P B; Queiros, S M Duarte; Jones, J L

2009-01-01

A metabolic model can be represented as a bipartite graph comprising linked reaction and metabolite nodes. Here it is shown how a network of conserved fluxes can be assigned to the edges of such a graph by combining the reaction fluxes with a conserved metabolite property such as molecular weight. A similar flux network can be constructed by combining the primal and dual solutions to the linear programming problem that typically arises in constraint-based modelling. Such constructions may help with the visualization of flux distributions in complex metabolic networks. The analysis also explains the strong correlation observed between metabolite shadow prices (the dual linear programming variables) and conserved metabolite properties. The methods were applied to recent metabolic models for Escherichia coli, Saccharomyces cerevisiae and Methanosarcina barkeri. Detailed results are reported for E. coli; similar results were found for other organisms

Modification of tokamak edge turbulence using feedback

International Nuclear Information System (INIS)

Richards, B.; Uckan, T.; Wootton, A.J.; Carreras, B.A.; Bengtson, R.D.; Hurwitz, P.; Li, G.X.; Lin, H.; Rowan, W.L.; Tsui, H.Y.W.; Sen, A.K.; Uglum, J.

1994-01-01

Using active feedback, the turbulent fluctuation levels have been reduced by as much as a factor of 2 in the edge of the Texas Experimental Tokamak (TEXT) [K. W. Gentle, Nucl. Fusion Technol. 1, 479 (1981)]. A probe system was used to drive a suppressor wave in the TEXT limiter shadow. A decrease in the local turbulence-induced particle flux has been seen, but a global change in the particle transport at the present time has not been observed. By changing the phase shift and gain of the feedback network, the amplitude of the turbulence was increased by a factor of 10

Safety factor profile control in a tokamak

CERN Document Server

Bribiesca Argomedo, Federico; Prieur, Christophe

2014-01-01

Control of the Safety Factor Profile in a Tokamak uses Lyapunov techniques to address a challenging problem for which even the simplest physically relevant models are represented by nonlinear, time-dependent, partial differential equations (PDEs). This is because of the  spatiotemporal dynamics of transport phenomena (magnetic flux, heat, densities, etc.) in the anisotropic plasma medium. Robustness considerations are ubiquitous in the analysis and control design since direct measurements on the magnetic flux are impossible (its estimation relies on virtual sensors) and large uncertainties remain in the coupling between the plasma particles and the radio-frequency waves (distributed inputs). The Brief begins with a presentation of the reference dynamical model and continues by developing a Lyapunov function for the discretized system (in a polytopic linear-parameter-varying formulation). The limitations of this finite-dimensional approach motivate new developments in the infinite-dimensional framework. The t...

Status of the tokamak program

Science.gov (United States)

Sheffield, J.

1981-08-01

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

The GBS code for tokamak scrape-off layer simulations

International Nuclear Information System (INIS)

Halpern, F.D.; Ricci, P.; Jolliet, S.; Loizu, J.; Morales, J.; Mosetto, A.; Musil, F.; Riva, F.; Tran, T.M.; Wersal, C.

2016-01-01

We describe a new version of GBS, a 3D global, flux-driven plasma turbulence code to simulate the turbulent dynamics in the tokamak scrape-off layer (SOL), superseding the code presented by Ricci et al. (2012) [14]. The present work is driven by the objective of studying SOL turbulent dynamics in medium size tokamaks and beyond with a high-fidelity physics model. We emphasize an intertwining framework of improved physics models and the computational improvements that allow them. The model extensions include neutral atom physics, finite ion temperature, the addition of a closed field line region, and a non-Boussinesq treatment of the polarization drift. GBS has been completely refactored with the introduction of a 3-D Cartesian communicator and a scalable parallel multigrid solver. We report dramatically enhanced parallel scalability, with the possibility of treating electromagnetic fluctuations very efficiently. The method of manufactured solutions as a verification process has been carried out for this new code version, demonstrating the correct implementation of the physical model.

Statistical properties of turbulent transport and fluctuations in tokamak and stellarator devices

Energy Technology Data Exchange (ETDEWEB)

Hidalgo, C; Pedrosa, M A; Milligen, B Van; Sanchez, E; Balbin, R; Garcia-Cortes, I [Euratom-CIEMAT Association, Madrid (Spain); Bleuel, J; Giannone, L.; Niedermeyer, H [Euratom-IPP Association, Garching (Germany)

1997-05-01

The statistical properties of fluctuations and turbulent transport have been studied in the plasma boundary region of stellarator (TJ-IU, W7-AS) and tokamak (TJ-I) devices. The local flux probability distribution function shows the bursty character of the flux and presents a systematic change as a function of the radial location. There exist large amplitude transport bursts that account for a significant part of the total flux. There is a strong similarity between the statistical properties of the turbulent fluxes in different devices. The value of the radial coherence associated with fluctuations and turbulent transport is strongly intermittent. This result emphasizes the importance of measurements with time resolution in understanding the interplay between the edge and the core regions in the plasma. For measurements in the plasma edge region of the TJ-IU torsatron, the turbulent flux does not, in general, show a larger radial coherence than the one associated with the fluctuations. (author). 14 refs, 6 figs.

Present status of Tokamak research

International Nuclear Information System (INIS)

Basu, Jayanta

1991-01-01

The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

Conservation laws with non-convex flux and applications to two-phase flow in porous media

Energy Technology Data Exchange (ETDEWEB)

Tegnander, Cathrine

1998-12-31

This thesis deals with conservation laws, which form a family of partial differential equations (PDEs) describing conservation of mass, momentum and energy. The first part studies some theoretical aspects of conservation laws: (1) Scalar hyperbolic conservation laws with a non-convex flux function, where time dependent decay estimates are mainly obtained by a front tracking technique, (2) Convergence of solutions for a finite difference scheme given by a class of one dimensional parabolic systems. The second part of the thesis applies the theory to multiphase flow in porous media. A number of mathematical models for multiphase flow in groundwater are studied. Techniques to improve the study of simulations of oil, gas and water phases in reservoirs such as in the North Sea are discussed. Upscaling of a refinement of the permeability field is evaluated using a flow simulation. This is done by a study of the preserving of the rank of a number of realizations with respect to the cumulative production parameter. Finally, the importance of selection of numerical methods in the simulations are exemplified by considering various splitting techniques. The numerical methods of front tracking and finite difference schemes and finite element methods are used. 98 refs., 24 figs., 18 tabs.

The scrape-off layer of a tokamak during the thermal phase of disruption

International Nuclear Information System (INIS)

Konkashbaev, I.K.

1993-01-01

The physical processes taking place in the scrape-off layer of a tokamak with a poloidal diverter during disruption are considered. It is shown that the physical processes in the scrape-off layer during disruption differ qualitatively from those in steady state. The main difference is that the plasma parameters in the scrape-off layer changes so as to facilitate transport along the field lines to the diverter plates, increasing the energy flux through the separatrix to disruption by a factor of 10 4 . It is found that for this the plasma in the scrape-off layer must already be hot and collisionless. During the transit time hot ions from the tokamak reach the diverter plates with essentially no energy loss. Because the electron velocity is large, an oppositely directed flux the wall plasma can be treated as infinite, i.e., electron recycling occurs. The energy lost to the scrape-off layer by anomalous thermal conductivity (diffusion) is transferred through turbulence to this cold electron stream by means of the two-stream instability. The mean electron energy ≅ 1 keV is substantially greater than that is steady state, T e ≅ 50 eV. Thus, an ion flux with E i ≅ 10 keV and a collisionless gas with T e ≅ 1 keV interact with the diverter plates. 3 refs., 4 figs

The Thomson Scattering System on the Lithium Tokamak eXperiment (LTX)

International Nuclear Information System (INIS)

Strickler, T.; Majeski, R.; Kaita, R.; LeBlanc, B.

2008-01-01

The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with R0 = 0.4m, a = 0.26m, BTF ∼ 3.4kG, IP ∼ 400kA, and pulse length ∼ 0.25s. The goal of LTX is to investigate tokamak plasmas that are almost entirely surrounded by a lithium-coated plasma-facing shell conformal to the last closed magnetic flux surface. Based on previous experimental results and simulation, it is expected that the low-recycling liquid lithium surfaces will result in higher temperatures at the plasma edge, flatter overall temperature profiles, centrally-peaked density profiles, and an increased confinement time. To test these predictions, the electron temperature and density profiles in LTX will be measured by a multi-point Thomson scattering system (TVTS). Initially, TS measurements will be made at up to 12 simultaneous points between the plasma center and plasma edge. Later, high resolution edge measurements will be deployed to study the lithium edge physics in greater detail. Technical challenges to implementing the TS system included limited 'line of sight' access to the plasma due to the plasma-facing shell and problems associated with the presence of liquid lithium.

Some stress-related issues in tokamak fusion reactor first walls

International Nuclear Information System (INIS)

Majumdar, S.; Pai, B.; Ryder, R.H.

1987-01-01

Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

Tokamak engineering mechanics

CERN Document Server

Song, Yuntao; Du, Shijun

2013-01-01

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

Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

International Nuclear Information System (INIS)

Olynyk, G.M.; Hartwig, Z.S.; Whyte, D.G.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Mumgaard, R.T.; Podpaly, Y.A.

2012-01-01

Highlights: ► A new scaling for obtaining reactor similarity in the divertor of scaled tokamaks. ► Conceptual design for a tokamak (“Vulcan”) to implement this new scaling. ► Demountable superconducting coils and compact neutron shielding. ► Helium-cooled high-temperature vacuum vessel and first wall. ► High-field-side lower hybrid current drive for non-inductive operation. - Abstract: An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ≃ 1 MW m −2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ∼ R −2/7 . While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system

Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

International Nuclear Information System (INIS)

Chen Yiping; Liu Songlin

2010-01-01

Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

Effect of nonequipotentiality of magnetic surfaces on the transport processes in a tokamak

International Nuclear Information System (INIS)

Kovrizhnykh, L.M.; Shasharina, S.G.

1987-01-01

The effect of deflection of equipotential from the magnetic surfaces on the transport processes in a tokamak is considered. The values of radial and poloidal electric fields are determined self-consistently, particle and heat fluxes are calculated with regard to these fields. It is shown that in some cases the effect of the poloidal electric field on the current values is very substantial

Economic evaluation of fissile fuel production using resistive magnet tokamaks

International Nuclear Information System (INIS)

Doyle, J.C. Jr.

1985-06-01

The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U 3 O 8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U 3 O 8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

Status of tokamak research

International Nuclear Information System (INIS)

Rawls, J.M.

1979-10-01

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

Experimental investigation of turbulent transport at the edge of a tokamak plasma

International Nuclear Information System (INIS)

Fedorczak, N.

2010-01-01

This manuscript is devoted to the experimental investigation of particle transport in the edge region of the tokamak Tore Supra. The first part introduces the motivations linked to energy production, the principle of a magnetic confinement and the elements of physics essential to describe the dynamic of the plasma at the edge region. From data collected by a set of Langmuir probes and a fast visible imaging camera, we demonstrate that the particle transport is dominated by the convection of plasma filaments, structures elongated along magnetic field lines. They present a finite wave number, responsible for the high enhancement of the particle flux at the low field side of the tokamak. This leads to the generation of strong parallel flows, and the strong constraint of filament geometry by the magnetic shear. (author)

Tokamak ARC damage

International Nuclear Information System (INIS)

Murray, J.G.; Gorker, G.E.

1985-01-01

Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

Murray, J.G.; Gorker, G.E.

1985-01-01

Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

Electrostatic turbulence in the Tokamak TBR-1

International Nuclear Information System (INIS)

Castro, R.M. de.

1991-01-01

Characteristics of turbulence at plasma edge of tokamak TBR - 1 are determined from measurements of potentials and density fluctuations, done with a square array of four single Langmuir probes. Two adjacent probes are used to measure the floating potential of the plasma in either poloidal or toroidal directions, the remaining two probes are used to measure saturation current also in poloidal and toroidal directions. Using multiple shot data from the four probe array the radial fluctuation density (n ∼ ) and floating potential (φ ∼ ) profiles are estimated. Analysing the fluctuations spectra the wavenumber-frequency spectrum S(k,ω) from two points measurements is determined. An extension of the cross-correlation concept to a three points correlations leads to the estimation of the fluctuation induced particle flux, from which the particle diffusion coefficient and the convected heat flux can be estimated. All this measurements were performed with and without a resonant magnetic field to verify the eventual influence of this field on the data already mentioned. It was verified that the particle flux is outward and due to electrostatic fluctuations with frequencies lower than 150 khz. (author)

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)

Anomalous transport in the tokamak edge

International Nuclear Information System (INIS)

Vayakis, G.

1991-04-01

The tokamak edge has been studied with arrays of Langmuir and magnetic probes on the DITE and COMPASS-C devices. Measurements of plasma parameters such as density, temperature and radial magnetic field were taken in order to elucidate the character, effect on transport and origin of edge fluctuations. The tokamak edge is a strongly-turbulent environment, with large electrostatic fluctuation levels and broad spectra. The observations, including direct correlation measurements, are consistent with a picture in which the observed magnetic field fluctuations are driven by the perturbations in electrostatic parameters. The propagation characteristics of the turbulence, investigated using digital spectral techniques, appear to be dominated by the variation of the radial electric field, both in limiter and divertor plasmas. A shear layer is formed, associated in each case with the last closed flux surface. In the shear layer, the electrostatic wavenumber spectra are significantly broader. The predictions of a drift wave model (DDGDT) and of a family of models evolving from the rippling mode (RGDT group), are compared with experimental results. RGDT, augmented by impurity radiation effects, is shown to be the most reasonable candidate to explain the nature of the edge turbulence, only failing in its estimate of the wavenumber range. (Author)

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Critique of atomic physics instability mechanisms: Ionization-driven and radiative microinstabilities in the tokamak edge plasma

International Nuclear Information System (INIS)

Ross, D.W.

1994-01-01

The theory of atomic-process driven microinstabilities in the tokamak edge plasma is reexamined. It is found that these instabilities, as they are usually presented, do not exist. This assertion applies both to ionization-driven modes and to radiative condensation, or thermal-driven modes. The problem is that there exists no separation of time scales between the approach to equilibrium and the growth rate of the purported instabilities. Therefore, to describe the perturbation of an inhomogeneous plasma, it is essential either to establish an equilibrium that includes both perpendicular transport and the proposed source, or, alternatively, to follow the background evolution simultaneously with the growth of the modes. Neither has been done in theoretical or numerical studies of microinstabilities driven by atomic effects in tokamaks. Very near the density limit, macroscopic modes may be unstable, leading to marfes or disruptions, but perturbations of the equilibrium transport fluxes, when taken into account, are sufficient to stabilize the microscopic modes. If the equilibrium fluxes are not included a priori, the ordering breakdown persists into the nonlinear regime. Since the atomic driving terms are the same as in the linear limit, radial decorrelation lengths would have to approach background scale lengths to yield transport of significant magnitude. Under ordinary tokamak conditions, therefore, atomic processes are unlikely to provide an important driving mechanism for the microturbulence that is presumed to cause anomalous transport

Poloidal density variation of impurities in a rotating tokamak plasma - flux surface coordinates and effect on transport coefficients

International Nuclear Information System (INIS)

Romanelli, M.

1999-09-01

The poloidal variation of impurity densities over magnetic surfaces brings about an enhancement of neoclassical transport coefficients, as shown by Romanelli and Ottaviani for impurities in the Pfirsch Schlueter regime and by Helander for particles in the banana-plateau regime, both in a large aspect ratio tokamak. The same effect will occur in a finite aspect ratio tokamak and therefore it is considered to be relevant for inclusion in transport codes for comparison with the experimental measurements of impurity transport. Here an expression for the impurity-density poloidal-variation generated by the fast toroidal rotation of the plasma column is presented in general coordinates. (author)

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)

ASDEX upgrade - definition of a tokamak experiment with a reactor compatible polaoidal divertor

International Nuclear Information System (INIS)

1982-03-01

ASDEX Upgrade is intended as the next experimental step after ASDEX. It is designed to investigate the physics of a divertor tokamak as closely as possible to fusion reactor requirements, without thermonuclear heating. It is characterized by a poloidal divertor configuration with divertor coils located outside the toroidal field coils, by machine parameters which allow a line density within the plasma boundary sufficient to screen fast CX particles from the plasma core, by a scrape-off layer essentially opaque to neutrals produced at the target plates, and, finally, by an auxiliary heating power high enough for producing a reactor-like power flux density through the plasma boundary. Design considerations on the basis of physical and technical constraints yielded the tokamak system optimized with respect to effort and costs as described in the following. It uses normal-conducting coil systems, is the size of ASDEX, and has a field of 3.9 T, a plasma current of up to 1.5 MA, and a pulse duration of 10 s. To provide the required power flux density, an ICRH power of 10 MW is needed. For comparison, a superconducting version is under investigation. (orig.)

Observation of drift wave propagation as a source of tokamak edge turbulence

International Nuclear Information System (INIS)

Wang Guiding; Liu Wandong; Yu Changxuan

1998-01-01

Core and edge turbulences were measured by Langmuir probe arrays in the KT-5C tokamak plasma. The radial wavenumber spectra show a quasimode like structure which results in a net radial outward propagation of the turbulent fluctuations. The measured fluctuation levels and wave action fluxes are in good agreement with model predictions by Mattor et al., suggesting that drift wave propagation could be a source of edge turbulence

ELMy-H mode as limit cycle and chaotic oscillations in tokamak plasmas

International Nuclear Information System (INIS)

Itoh Sanae, I.; Itoh, Kimitaka; Fukuyama, Atsushi; Miura, Yukitoshi.

1991-05-01

A model of Edge Localized Modes (ELMs) in tokamak plasmas is presented. A limit cycle solution is found in the transport equation (time-dependent Ginzburg-Landau type), which a has hysteresis curve between the gradient and flux. Periodic oscillation of the particle outflux and L/H intermediate state are predicted near the L/H transition boundary. A mesophase in spatial structure appears near edge. Chaotic oscillation is also predicted. (author)

Tokamak confinement scaling laws

International Nuclear Information System (INIS)

Connor, J.

1998-01-01

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

Ballistic propagation of turbulence front in tokamak edge plasmas

International Nuclear Information System (INIS)

Sugita, Satoru; Itoh, Kimitaka; Itoh, Sanae-I; Yagi, Masatoshi; Fuhr, Guillaume; Beyer, Peter; Benkadda, Sadruddin

2012-01-01

The flux-driven nonlinear simulation of resistive ballooning mode turbulence with tokamak edge geometry is performed to study the non-steady component in the edge turbulence. The large-scale and dynamical events in transport are investigated in a situation where the mean flow is suppressed. Two types of dynamics are observed. One is the radial propagation of the pulse of pressure gradient, the other is the appearance/disappearance of radially elongated global structure of turbulent heat flux. The ballistic propagation is observed in the pulse of pressure gradient, which is associated with the front of turbulent heat flux. We focus on this ballistic propagation phenomenon. Both of the bump of pressure gradient and the front of heat flux propagate inward and outward direction. It is confirmed that the strong fluctuation propagates with the pulse front. It is observed that the number of pulses going outward is close to those going inward. This ballistic phenomenon does not contradict to the turbulence spreading theory. Statistical characteristics of the ballistic propagation of pulses are evaluated and compared with scaling laws which is given by the turbulence spreading theory. It is found that they give qualitatively good agreement. (paper)

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

Large Aspect Ratio Tokamak Study

International Nuclear Information System (INIS)

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

1980-06-01

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

Influence of the plasma edge on tokamak performance

International Nuclear Information System (INIS)

Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

2000-01-01

A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge MHD instabilities and plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. The article examines these phenomena and their interaction. (author)

Influence of the plasma edge on tokamak performance

International Nuclear Information System (INIS)

Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

1999-01-01

A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

Influence of the plasma edge on tokamak performance

International Nuclear Information System (INIS)

Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

2001-01-01

A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

The reconstruction and research progress of the TEXT-U tokamak in China

Science.gov (United States)

Zhuang, G.; Pan, Y.; Hu, X. W.; Wang, Z. J.; Ding, Y. H.; Zhang, M.; Gao, L.; Zhang, X. Q.; Yang, Z. J.; Yu, K. X.; Gentle, K. W.; Huang, H.; J-TEXT Team

2011-09-01

The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 × 1019 m-3, and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.

The reconstruction and research progress of the TEXT-U tokamak in China

International Nuclear Information System (INIS)

Zhuang, G.; Pan, Y.; Hu, X.W.; Wang, Z.J.; Ding, Y.H.; Zhang, M.; Gao, L.; Zhang, X.Q.; Yang, Z.J.; Yu, K.X.; Gentle, K.W.; Huang, H.

2011-01-01

The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 x 10 19 m -3 , and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.

Self-similar density turbulence in the TCV tokamak scrape-off layer

International Nuclear Information System (INIS)

Graves, J P; Horacek, J; Pitts, R A; Hopcraft, K I

2005-01-01

Plasma fluctuations in the scrape-off layer (SOL) of the TCV tokamak exhibit statistical properties which are universal across a broad range of discharge conditions. Electron density fluctuations, from just inside the magnetic separatrix to the plasma-wall interface, are described well by a gamma distributed random variable. The density fluctuations exhibit clear evidence of self-similarity in the far SOL, such that the corresponding probability density functions collapse upon renormalization solely by the mean particle density. This constitutes a demonstration that the amplitude of the density fluctuations is simply proportional to the mean density and is consistent with the further observation that the radial particle flux fluctuations scale solely with the mean density over two orders of magnitude. Such findings indicate that it may be possible to improve the prediction of transport in the critical plasma-wall interaction region of future large scale tokamaks. (letter to the editor)

Developments in tokamak transport modeling

International Nuclear Information System (INIS)

Houlberg, W.A.; Attenberger; Lao, L.L.

1981-01-01

A variety of numerical methods for solving the time-dependent fluid transport equations for tokamak plasmas is presented. Among the problems discussed are techniques for solving the sometimes very stiff parabolic equations for particle and energy flow, treating convection-dominated energy transport that leads to large cell Reynolds numbers, optimizing the flow of a code to reduce the time spent updating the particle and energy source terms, coupling the one-dimensional (1-D) flux-surface-averaged fluid transport equations to solutions of the 2-D Grad-Shafranov equation for the plasma geometry, handling extremely fast transient problems such as internal MHD disruptions and pellet injection, and processing the output to summarize the physics parameters over the potential operating regime for reactors. Emphasis is placed on computational efficiency in both computer time and storage requirements

Poloidal inhomogeneity of the particle fluctuation induced fluxes near of the LCFS at lower hybrid heating and improved confinement transition at the FT-2 tokamak

International Nuclear Information System (INIS)

Lashkul, S.I.; Altukhov, A.B.; Gurchenko, A.D.; Gusakov, E.Z.; Dyachenko, V.V.; Esipov, L.A.; Kantor, M.Y.; Kouprienko, D.V.; Stepanov, A.Y.; Sharpeonok, A.P.; Shatalin, S.V.; Vekshina, E.O.

2004-01-01

This paper present our observations and conclusions about development of the transport process at the plasma periphery of the small tokamak FT-2 during additional Lower Hybrid Heating (LHH), when external (ETB) transport barrier followed by Internal (ITB) transport barrier is observed. The peculiarities of the variations of the fluctuation fluxes near periphery are measured by three moveable multi-electrode Langmuir probes (L-probe) located in the same poloidal cross-section of the chamber. So the observed L-H transition and ETB formation after LHH and the associated negative E r rise result mainly from the decrease of the electron temperature (T e ) near inner region of the LCFS (last close flux surface) by greater extent than in SOL (scrape-off layer). This effect is stimulated by decrease of the input power and decrease of the radial correlation coefficient (for r equals 74-77 mm) (and radial particle fluctuation-induced Γ(t)) resulted from ITB formation mechanism during LHH. T e variation in the SOL after LH heating pulse takes place to a lesser extent. Observed non-monotonic radial profile of T e near LCFS with positive δT e /δr rise is kept constant obviously by large longitudinal conductivity and poloidal fluxes from the hotter limiter shadow regions because of the poloidal inhomogeneity of the T e (SOL) and n e (SOL). Such induced negative E r after RF pulse gives fast rise to a quasi-steady-state Γ 0 (t) drift fluxes with reversed direction structure, like 'zonal flows', which may inhibit transport across the flow. Large rise of grad(n e ) after LHH near LCFS with L-H transition is observed after the end of LH pulse for a long time - about 10-15 ms

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

International Nuclear Information System (INIS)

Shimada, Ryuichi

1982-01-01

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

Fusion potential for spherical and compact tokamaks

International Nuclear Information System (INIS)

Sandzelius, Mikael

2003-02-01

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

Fusion potential for spherical and compact tokamaks

Energy Technology Data Exchange (ETDEWEB)

Sandzelius, Mikael

2003-02-01

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

HOMOCLINIC TANGLE BIFURCATIONS AND EDGE STOCHASTICITY IN DIVERTED TOKAMAKS

International Nuclear Information System (INIS)

EVANS, T.E.; ROEDER, R.K.W.; CARTER, J.A.; RAPOPORT, B.I.

2003-01-01

OAK-B135 The boundary and pedestal region of a poloidally diverted tokamak is particularly susceptible to the onset of vacuum magnetic field stochasticity due to small non-axisymmetric resonant perturbations. Recent calculations of the separatrix topology in diverted tokamaks, when subjected to small magnetic perturbations, show the existence of complex invariant manifold structures known as homoclinic tangles. These structures appear above a relatively low perturbation threshold that depends on certain equilibrium shape parameters. Homoclinic tangles represent a splitting of the unperturbed separatrix into stable and unstable invariant manifolds associated with each X-point (hyperbolic point). The manifolds that make up homoclinic tangles set the boundaries that prescribe how stochastic field line trajectories are organized i.e., how field lines from the inner domain of the unperturbed separatrix mix and are transported to plasma facing surfaces such as divertor target plates and protruding baffle structures. Thus, the topology of these tangles determines which plasma facing components are most likely to interact with escaping magnetic field lines and the parallel heat and particle flux they carry

Scaling of divertor heat flux profile widths in DIII-D

International Nuclear Information System (INIS)

Lasnier, C.J.; Makowski, M.A.; Boedo, J.A.; Allen, S.L.; Brooks, N.H.; Hill, D.N.; Leonard, A.W.; Watkins, J.G.; West, W.P.

2011-01-01

New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

Microwave Tokamak Experiment

International Nuclear Information System (INIS)

Anon.

1988-01-01

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

Magnetohydrodynamic helical structures in nominally axisymmetric low-shear tokamak plasmas

International Nuclear Information System (INIS)

Graves, J P; Brunetti, D; Cooper, W A; Reimerdes, H; Halpern, F; Pochelon, A; Sauter, O; Chapman, I T

2013-01-01

The primary goal of hybrid scenarios in tokamaks is to enable high performance operation with large plasma currents whilst avoiding MHD instabilities. However, if a local minimum in the safety factor is allowed to approach unity, the energy required to overcome stabilizing magnetic field line bending is very small, and as a consequence, large MHD structures can be created, with typically dominant m = n = 1 helical component. If there is no exact q = 1 rational surface the essential character of these modes can be modelled assuming ideal nested magnetic flux surfaces. The methods used to characterize these structures include linear and non-linear ideal MHD stability calculations which evaluate the departure from an axisymmetric plasma state, and also equilibrium calculations using a 3D equilibrium code. While these approaches agree favourably for simulations of ITER relevant hybrid regimes in this paper, the relevance of the ideal MHD model itself is tested through empirical examination of helical states in MAST and TCV. While long lived modes in MAST do not have island structures, some of the continuous mode oscillations exhibited in high elongation experiments in TCV indicate that resistivity may play a role in further weakening the ability of the tokamak core to remain axisymmetric. The simulations and experiments consistently highlight the need to control the safety factor in hybrid scenarios planned for future fusion grade tokamaks such as ITER. (paper)

Morphodynamic Modeling of the Lower Yellow River, China: Flux (Equilibrium) Form or Entrainment (Nonequilibrium) Form of Sediment Mass Conservation?

Science.gov (United States)

An, C.; Parker, G.; Ma, H.; Naito, K.; Moodie, A. J.; Fu, X.

2017-12-01

Models of river morphodynamics consist of three elements: (1) a treatment of flow hydraulics, (2) a formulation relating some aspect of sediment transport to flow hydraulics, and (3) a description of sediment conservation. In the case of unidirectional river flow, the Exner equation of sediment conservation is commonly described in terms of a flux-based formulation, in which bed elevation variation is related to the streamwise gradient of sediment transport rate. An alternate formulation of the Exner equation, however, is the entrainment-based formulation in which bed elevation variation is related to the difference between the entrainment rate of bed sediment into suspension and the deposition rate of suspended sediment onto the bed. In the flux-based formulation, sediment transport is regarded to be in a local equilibrium state (i.e., sediment transport rate locally equals sediment transport capacity). However, the entrainment-based formulation does not require this constraint; the sediment transport rate may lag in space and time behind the changing flow conditions. In modeling the fine-grained Lower Yellow River, it is usual to treat sediment conservation in terms of an entrainment-based (nonequilibrium) rather than a flux-based (equilibrium) formulation with the consideration that fine-grained sediment may be entrained at one place but deposited only at some distant location downstream. However, the differences in prediction between the two formulations are still not well known, and the entrainment formulation may not always be necessary for the Lower Yellow River. Here we study this problem by comparing the results of flux-based and entrainment-based morphodynamics under conditions typical of the Yellow River, using sediment transport equations specifically designed for the Lower Yellow River. We find, somewhat unexpectedly, that in a treatment of a 200-km reach using uniform sediment, there is little difference between the two formulations unless the

Study of plasma discharge evolution and edge turbulence with fast visible imaging in the Aditya tokamak

International Nuclear Information System (INIS)

Banerjee, Santanu; Manchanda, R.; Chowdhuri, M.B.

2015-01-01

Study of discharge evolution through the different phases of a tokamak plasma shot viz., the discharge initiation, current ramp-up, current flat-top and discharge termination, is essential to address many inherent issues of the operation of a Tokamak. Fast visible imaging of the tokamak plasma can provide valuable insight in this regard. Further, edge turbulence is considered to be one of the quintessential areas of tokamak research as the edge plasma is at the immediate vicinity of the plasma core and plays vital role in the core plasma confinement. The edge plasma also bridges the core and the scrape off layer (SOL) of the tokamak and hence has a bearing on the particle and heat flux escaping the plasma column. Two fast visible imaging systems are installed on the Aditya tokamak. One of the system is for imaging the plasma evolution with a wide angle lens covering a major portion of the vacuum vessel. The imaging fiber bundle along with the objective lens is installed inside a radial re-entrant viewport, specially designed for the purpose. Another system is intended for tangential imaging of the plasma column. Formation of the plasma column and its evolution are studied with the fast visible imaging in Aditya. Features of the ECRH and LHCD operations on Aditya will be discussed. 3D filaments can, be seen at the plasma edge all along the discharge and they get amplified in intensity at the plasma termination phase. Statistical analysis of these filaments, which are essentially plasma blobs will be presented. (author)

Measurement of inner wall limiter SOL widths in KSTAR tokamak

Directory of Open Access Journals (Sweden)

J.G. Bak

2017-08-01

Full Text Available Scrape-off layer (SOL widths λq are presented from the KSTAR tokamak using fast reciprocating Langmuir probe assembly (FRLPA measurements at the outboard mid-plane (OMP and the infra-Red (IR thermography at inboard limiter tiles in moderately elongated (κ = 1.45 – 1.55 L-mode inner wall-limited (IWL plasmas under experimental conditions such as BT = 2.0 T, PNBI = 1.4 – 1.5 MW, line averaged densities 2.5 – 5.1 × 1019 m−3 and plasma current Ip = 0.4 − 0.7 MA. There is clear evidence for a double exponential structure in q||(r from the FRLPA such that, for example at Ip = 0.6 MA, a narrow feature, λq,near (=3.5 mm is found close to the LFCS, followed by a broader width, λq,main (=57.0 mm. Double exponential profiles (λq,near = 1.5 – 2.8 mm, λq,main = 17.0 – 35.0 mm can be also observed in the IR heat flux mapped to the OMP throughout the range of Ip investigated. In addition, analysis of SOL turbulence statistics obtained with the FRLPA shows high relative fluctuation levels and positively skewed distributions in electron temperature and ion particle flux across the SOL, with both properties increasing for longer distance from the LCFS, as often previously observed in the tokamaks. Interestingly, the fluctuation character expressed in terms of spectral distributions remains unchanged in passing from the narrow to the broad SOL heat flux channel.

Overview of recent experimental results from the DIII-D advanced tokamak program

International Nuclear Information System (INIS)

Burrell, K.H.

2003-01-01

The D III-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: 1) We have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; 2) Using this rotational stabilization, we have achieved β N H 89 ≥ 10 for 4 τ E limited by the neoclassical tearing mode; 3) Using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased β T by 60%; 4) We have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; 5) We have made the first integrated AT demonstration discharges with current profile control using ECCD; 6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and 7) We have demonstrated stationary tokamak operation for 6.5 s (36 τ E ) at the same fusion gain parameter of β N H 89 /q 95 2 ≅ 0.4 as ITER but at much higher q 95 = 4.2. We have developed general improvements applicable to conventional and advanced tokamak operating modes: 1) We have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 τ E ) with constant density and constant radiated power; 2) We have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; 3) We have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. (author)

Reconstruction of vacuum magnetic flux in QUEST

International Nuclear Information System (INIS)

Ishiguro, Masaki; Hanada, Kazuaki; Nakamura, Kazuo

2010-01-01

It is important to determine the best method for reconstructing the magnetic flux when eddy currents are significantly induced during magnetic measurement in spherical tokamaks (STs). Four methods for this reconstruction are investigated, and the calculated magnetic fluxes are compared to those measured in the cavity of a vacuum vessel. The results show that the best method is the one that uses currents from virtual coils for reconstruction. In this method, the placement of the virtual coils is optimized with numerical simulations using the Akaike information criterion (AIC), which indicates the goodness of fit of models used to fit measured data. The virtual coils are set on a line 15 cm outside the vacuum vessel. (author)

Performance of V-4Cr-4Ti Alloy Exposed to the JFT-2M Tokamak Environment

International Nuclear Information System (INIS)

Johnson, W.R.; Trester, P.W.; Sengoku, S.; Ishiyama, S.; Fukaya, K.; Eto, M.; Oda, T.; Hirohata, Y.; Hino, T.; Tsai, H.

1999-01-01

A long-term test has been conducted in the JFT-2M tokamak fusion device to determine the effects of environmental exposure on the mechanical and chemical behavior of a V-4Cr-4Ti alloy. Test specimens of the alloy were exposed in the outward lower divertor chamber of JFT-2M in a region away from direct contact with the plasma and were preheated to 300 C just prior to and during selected plasma discharges. During their nine-month residence time in JFT-2M, the specimens experienced approximately 200 lower single-null divertor shots at 300 C, during which high energy particle fluxes to the preheated test specimens were significant, and approximately 2,010 upper single-null divertor shots and non-diverter shots at room temperature, for which high energy particle fluxes to and expected particle retention in the test specimens were very low. Data from post-exposure tests have indicated that the performance of the V-4Cr-4Ti alloy would not be significantly affected by environmental exposure to gaseous species at partial pressures typical for tokamak operation. Deuterium retention in the exposed alloy was also low (<2 ppm). Absorption of interstitial by the alloy was limited to the very near surface, and neither the strength nor the Charpy impact properties of the alloy appeared to be significantly changed from the exposure to the JFT-2M tokamak environment

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.

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)

Tokamak control simulator

International Nuclear Information System (INIS)

Edelbaum, T.N.; Serben, S.; Var, R.E.

1976-01-01

A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

Particle control and plasma performance in the Lithium Tokamak eXperiment

Energy Technology Data Exchange (ETDEWEB)

Majeski, R.; Abrams, T.; Boyle, D.; Granstedt, E.; Hare, J.; Jacobson, C. M.; Kaita, R.; Kozub, T.; LeBlanc, B.; Lundberg, D. P.; Lucia, M.; Merino, E.; Schmitt, J.; Stotler, D. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Biewer, T. M.; Canik, J. M.; Gray, T. K.; Maingi, R.; McLean, A. G. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Kubota, S. [University of California at Los Angeles, Los Angeles, California 90095 (United States); and others

2013-05-15

The Lithium Tokamak eXperiment is a small, low aspect ratio tokamak [Majeski et al., Nucl. Fusion 49, 055014 (2009)], which is fitted with a stainless steel-clad copper liner, conformal to the last closed flux surface. The liner can be heated to 350 °C. Several gas fueling systems, including supersonic gas injection and molecular cluster injection, have been studied and produce fueling efficiencies up to 35%. Discharges are strongly affected by wall conditioning. Discharges without lithium wall coatings are limited to plasma currents of order 10 kA, and discharge durations of order 5 ms. With solid lithium coatings discharge currents exceed 70 kA, and discharge durations exceed 30 ms. Heating the lithium wall coating, however, results in a prompt degradation of the discharge, at the melting point of lithium. These results suggest that the simplest approach to implementing liquid lithium walls in a tokamak—thin, evaporated, liquefied coatings of lithium—does not produce an adequately clean surface.

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}.

Equilibrium and ballooning mode stability of an axisymmetric tensor pressure tokamak

International Nuclear Information System (INIS)

Cooper, W.A.; Bateman, G.; Nelson, D.B.; Kammash, T.

1980-08-01

A force balance relation, a representation for the poloidal beta (β/sub p/), and expressions for the current densities are derived from the MHD equilibrium relations for an axisymmetric tensor pressure tokamak. Perpendicular and parallel beam pressure components are evaluated from a distribution function that models high energy neutral particle injection. A double adiabatic energy principle is derived from that of Kruskal and Oberman, with correction terms added. The energy principle is then applied to an arbitrary cross-section axisymmetric tokamak to examine ballooning instabilities of large toroidal mode number. The resulting Euler equation is remarkably similar to that of ideal MHD. Although the field-bending term is virtually unaltered, the driving term is modified because the pressures are no longer constant on a flux surface. Either a necessary or a sufficient marginal stability criterion for a guiding center plasma can be derived from this equation whenever an additional stabilizing element unique to the double adiabatic theory is either kept or neglected, respectively

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

Hybrid model for simulation of plasma jet injection in tokamak

Science.gov (United States)

Galkin, Sergei A.; Bogatu, I. N.

2016-10-01

Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

Survey of Tokamak experiments

International Nuclear Information System (INIS)

Bickerton, R.J.

1977-01-01

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

ELMy-H mode as limit cycle and chaotic oscillations in tokamak plasmas

International Nuclear Information System (INIS)

Itoh, Sanae; Itoh, Kimitaka; Fukuyama, Atsushi.

1991-06-01

A model of Edge Localized Modes (ELMs) in tokamaks is presented. A limit cycle solution is found in time-dependent Ginzburg Landau type model equation of L/H transition, which has a hysteresis curve between the plasma gradient and flux. The oscillation of edge density appears near the L/H transition boundary. Spatial structure of the intermediate state (mesophase) is obtained in the edge region. Chaotic oscillation is predicted due to random neutrals and external oscillations. (author)

Tokamak building-design considerations for a large tokamak device

International Nuclear Information System (INIS)

Barrett, R.J.; Thomson, S.L.

1981-01-01

Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

Real-time diamagnetic flux measurements on ASDEX Upgrade.

Science.gov (United States)

Giannone, L; Geiger, B; Bilato, R; Maraschek, M; Odstrčil, T; Fischer, R; Fuchs, J C; McCarthy, P J; Mertens, V; Schuhbeck, K H

2016-05-01

Real-time diamagnetic flux measurements are now available on ASDEX Upgrade. In contrast to the majority of diamagnetic flux measurements on other tokamaks, no analog summation of signals is necessary for measuring the change in toroidal flux or for removing contributions arising from unwanted coupling to the plasma and poloidal field coil currents. To achieve the highest possible sensitivity, the diamagnetic measurement and compensation coil integrators are triggered shortly before plasma initiation when the toroidal field coil current is close to its maximum. In this way, the integration time can be chosen to measure only the small changes in flux due to the presence of plasma. Two identical plasma discharges with positive and negative magnetic field have shown that the alignment error with respect to the plasma current is negligible. The measured diamagnetic flux is compared to that predicted by TRANSP simulations. The poloidal beta inferred from the diamagnetic flux measurement is compared to the values calculated from magnetic equilibrium reconstruction codes. The diamagnetic flux measurement and TRANSP simulation can be used together to estimate the coupled power in discharges with dominant ion cyclotron resonance heating.

Internal transport barrier triggering by rational magnetic flux surfaces in tokamaks

International Nuclear Information System (INIS)

Joffrin, E.; Challis, C.D.; Conway, G.D.

2003-01-01

The formation of Internal Transport Barriers (ITBs) has been experimentally associated with the presence of rational q-surfaces in both JET and ASDEX Upgrade. The triggering mechanisms are related to the occurrence of magneto-hydrodynamic (MHD) instabilities such as mode coupling or fishbone activity. These events could locally modify the poloidal velocity and increase transiently the shearing rate to values comparable to the linear growth rate of ITG modes. For JET reversed magnetic shear scenarios, ITB emergence occurs preferentially when the minimum q reaches an integer value. In this case, transport effects localised in the vicinity of zero magnetic shear and close to rational q values may also contribute to the formation of ITBs.The role of rational q surfaces on ITB triggering stresses the importance of q profile control for advanced tokamak scenario and could contribute to lower substantially the access power to these scenarios in next step facilities. (author)

Internal Transport Barrier triggering by rational magnetic flux surfaces in tokamaks

International Nuclear Information System (INIS)

Joffrin, E.H.

2002-01-01

The formation of Internal Transport Barriers (ITBs) has been experimentally associated with the presence of rational q-surfaces in both JET and ASDEX Upgrade. The triggering mechanisms are related to the occurrence of magneto-hydrodynamic (MHD) instabilities such as mode coupling or fishbone activity. These events could locally modify the poloidal velocity and increase transiently the shearing rate to values comparable to the linear growth rate of ITG modes. For reversed magnetic shear scenario, ITB emergence occurs preferentially when the minimum q reaches an integer value. In this case, transport effects localised in the vicinity of zero magnetic shear and close to rational q values may also contribute to the formation of ITBs. The role of rational q surfaces on ITB triggering stresses the importance of q profile control for advanced tokamak scenario and could contribute to lower substantially the access power to these scenarios in next step facilities. (author)

Modeling of sawtooth destabilization during radio-frequency heating experiments in tokamak plasmas

International Nuclear Information System (INIS)

McClements, K.G.; Dendy, R.O.; Hastie, R.J.; Martin, T.J.

1996-01-01

Sawtooth oscillations in tokamaks have been stabilized using ion cyclotron resonance heating (ICRH), but often reappear while ICRH continues. It is shown that the reappearance of sawteeth during one particular ICRH discharge in the Joint European Torus (JET) [Campbell et al., Phys. Rev. Lett. 60, 2148 (1988)] was correlated with a change of sign in the energy δW associated with m=1 internal kink displacements. To compute δW, a new analytical model is used for the distribution function of heated minority ions, which is consistent with Fokker endash Planck simulations of ICRH. Minority ions have a stabilizing influence, arising from third adiabatic invariant conservation, but also contribute to a destabilizing shift of magnetic flux surfaces. As the minor radius of the q=1 surface rises, the stabilizing influence of minority ions diminishes, and the shape of the plasma cross section becomes increasingly important. It is shown that an increase in ICRH power can destabilize the kink mode: this is consistent with observations of sawteeth in JET discharges with varying levels of ICRH. It is suggested that the sawtooth-free period could be prolonged by minimizing the vertical extent of the ICRH power deposition profile.1996 American Institute of Physics

Classical tokamak transport theory

International Nuclear Information System (INIS)

Nocentini, Aldo

1982-01-01

A qualitative treatment of the classical transport theory of a magnetically confined, toroidal, axisymmetric, two-species plasma is presented. The 'weakly collisional' ('banana' and 'plateau') and 'collision dominated' ('Pfirsch-Schlueter' and 'highly collisional') regimes, as well as the Ware effect are discussed. The method used to evaluate the diffusion coffieicnts of particles and heat in the weakly collisional regime is based on stochastic argument, that requires an analysis of the characteristic collision frequencies and lengths for particles moving in a tokamak-like magnetic field. The same method is used to evaluate the Ware effect. In the collision dominated regime on the other hand, the particle and heat fluxes across the magnetic field lines are dominated by macroscopic effects so that, although it is possible to present them as diffusion (in fact, the fluxes turn out to be proportional to the density and temperature gradients), a macroscopic treatment is more appropriate. Hence, fluid equations are used to inveatigate the collision dominated regime, to which particular attention is devoted, having been shown relatively recently that it is more complicated than the usual Pfirsch-Schlueter regime. The whole analysis presented here is qualitative, aiming to point out the relevant physical mechanisms involved in the various regimes more than to develop a rigorous mathematical derivation of the diffusion coefficients, for which appropriate references are given. (author)

Development of numerical methods to calculate the propagation and the absorption of the hybrid wave in tokamaks; Developpement des methodes numeriques pour la resolution de la propagation et de l`absorption de l`onde hybride dans les tokamaks

Energy Technology Data Exchange (ETDEWEB)

Sebelin, E

1997-12-15

Full-wave calculations based on trial functions are carried out for solving the lower hybrid current drive problem in tokamaks. A variational method is developed and provides an efficient system to describe in a global manner both the propagation and the absorption of the electromagnetic waves in plasmas. The calculation is fully carried out in the case of circular and concentric flux surfaces. The existence and uniqueness of the solution of the wave propagation equation is mathematically proved. The first realistic simulations are performed for the high aspect ratio tokamak TRIAM-1M. It is checked that the main features of the lower-hybrid wave dynamics are well described numerically. (A.C.) 81 refs.

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

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

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

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

1983-01-01

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

Non-existence of Normal Tokamak Equilibria with Negative Central Current

International Nuclear Information System (INIS)

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

2003-01-01

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

Development of numerical methods to calculate the propagation and the absorption of the hybrid wave in tokamaks

International Nuclear Information System (INIS)

Sebelin, E.

1997-01-01

Full-wave calculations based on trial functions are carried out for solving the lower hybrid current drive problem in tokamaks. A variational method is developed and provides an efficient system to describe in a global manner both the propagation and the absorption of the electromagnetic waves in plasmas. The calculation is fully carried out in the case of circular and concentric flux surfaces. The existence and uniqueness of the solution of the wave propagation equation is mathematically proved. The first realistic simulations are performed for the high aspect ratio tokamak TRIAM-1M. It is checked that the main features of the lower-hybrid wave dynamics are well described numerically. (A.C.)

COMPREHENSIVE GYROKINETIC SIMULATION OF TOKAMAK TURBULENCE AT FINITE RELATIVE GYRORADIUS

International Nuclear Information System (INIS)

WALTZ, R.E.; CANDY, J.; ROSENBLUTH, M.N.

2002-01-01

OAK B202 COMPREHENSIVE GYROKINETIC SIMULATION OF TOKAMAK TURBULENCE AT FINITE RELATIVE GYRORADIUS. A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate turbulent transport in actual experimental profiles and allow direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite beta, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius (ρ*) so as to treat the profile shear stabilization effects which break gyroBohm scaling. The code operates in a cyclic flux tube limit which allows only gyroBohm scaling and a noncyclic radial annulus with physical profile variation. The later requires an adaptive source to maintain equilibrium profiles. Simple ITG simulations demonstrate the broken gyroBohm scaling depends on the actual rotational velocity shear rates competing with mode growth rates, direct comprehensive simulations of the DIII-D ρ*-scaled L-mode experiments are presented as a quantitative test of gyrokinetics and the paradigm

Impurity transport calculations for the limiter shadow region of a tokamak

International Nuclear Information System (INIS)

Claassen, H.A.; Repp, H.

1981-01-01

Impurity transport calculations are presented for the scrape-off layer of a tokamak with a poloidal ring limiter. The theory is based on the drift-kinetic equations for the impurity ions in their different ionization states. It is developed in the limit of low impurity concentrations under due consideration of electron impact ionization, Coulomb collisions with hydrogen ions streaming onto a neutralizing surface, a convection along the magnetic field, and a radial drift. The background plasma and the impurity sources at the walls enter the theory as input parameters. Numerical results are given for the radial profiles of density, temperature, particle flux, and energy flux of wall-released impurity ions as well as for the screening efficiency of the scrape-off layer neglecting impurity re-emission from the limiter. (author)

Feedback stabilization of the axisymmetric instability of a deformable tokamak plasma

International Nuclear Information System (INIS)

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

1989-01-01

The paper presents an analysis of the magnetohydrodynamic stability of the axisymmetric system consisting of a free boundary tokamak plasma with non-circular cross-section, finite resistivity passive conductors, and an active feedback system with magnetic flux pickup loops, a proportional amplifier with gain G and current carrying poloidal field coils. A numerical simulation of the system when G is set to zero identifies flux loop locations which correctly sense the plasma motion. However, when certain of these locations are incorporated into an active feedback scheme, the plasma fails to be stabilized, no matter what value of the gain is chosen. Analysis on the basis of an extended energy principle indicates that this failure is due to the deformability of the plasma cross-section. (author). 14 refs, 7 figs

Neural network evaluation of tokamak current profiles for real time control

Science.gov (United States)

Wróblewski, Dariusz

1997-02-01

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

Neural network evaluation of tokamak current profiles for real time control

International Nuclear Information System (INIS)

Wroblewski, D.

1997-01-01

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

Modelling of Power Fluxes during Thermal Quenches

International Nuclear Information System (INIS)

Konz, C.; Coster, D. P.; Lackner, K.; Pautasso, G.

2005-01-01

Plasma disruptions, i. e. the sudden loss of magnetic confinement, are unavoidable, at least occasionally, in present day and future tokamaks. The expected energy fluxes to the plasma facing components (PFCs) during disruptions in ITER lie in the range of tens of GW/m''2 for timescales of about a millisecond. Since high energy fluxes can cause severe damage to the PFCs, their design heavily depends on the spatial and temporal distribution of the energy fluxes during disruptions. We investigate the nature of power fluxes during the thermal quench phase of disruptions by means of numerical simulations with the B2 SOLPS fluid code. Based on an ASDEX Upgrade shot, steady-state pre-disruption equilibria are generated which are then subjected to a simulated thermal quench by artificially enhancing the perpendicular transport in the ion and electron channels. The enhanced transport coefficients flows the Rechester and Rosenbluth model (1978) for ergodic transport in a tokamak with destroyed flux surfaces, i. e. χ, D∼const. xT''5/2 where the constants differ by the square root of the mass ratio for ions and electrons. By varying the steady-state neutral puffing rate we can modify the divertor conditions in terms of plasma temperature and density. Our numerical findings indicate that the disruption characteristics depend on the pre disruptive divertor conditions. We study the timescales and the spatial distribution of the divertor power fluxes. The simulated disruptions show rise and decay timescales in the range observed at ASDEX Upgrade. The decay timescale for the central electron temperature of ∼800 μs is typical for non-ITB disruptions. Varying the divertor conditions we find a distinct transition from a regime with symmetric power fluxes to inboard and outboard divertors to a regime where the bulk of the power flux goes to the outboard divertor. This asymmetry in the divertor peak fluxes for the higher puffing case is accompanied by a time delay between the

Summary discussion: An integrated advanced tokamak reactor

International Nuclear Information System (INIS)

Sauthoff, N.R.

1994-01-01

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

Internal transport barrier triggering by rational magnetic flux surfaces in tokamaks

International Nuclear Information System (INIS)

Joffrin, E.; Challis, C.D.; Conway, G.D.

2003-01-01

The formation of internal transport barriers (ITBs) has been experimentally associated with the presence of rational q surfaces in both JET and ASDEX Upgrade. The triggering mechanisms are related to the occurrence of magneto-hydrodynamic (MHD) instabilities such as mode coupling and fishbone activity. These events could locally modify the poloidal velocity and increase transiently the shearing rate to values comparable with the linear growth rate of ion temperature gradient modes. For JET reversed magnetic shear scenarios, ITB emergence occurs preferentially when the minimum q reaches an integral value. In this case, transport effects localized in the vicinity of zero magnetic shear and close to rational q values may be at the origin of ITB formation. The role of rational q surfaces in ITB triggering stresses the importance of q profile control for an advanced tokamak scenario and could assist in substantially lowering the access power to these scenarios in next step facilities. (author)

Improved core-edge tokamak transport simulations with the CORSICA 2 code

International Nuclear Information System (INIS)

Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

1996-01-01

The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

Neoclassical Physics for Current Drive in Tokamak Plasmas

International Nuclear Information System (INIS)

Duthoit, F.X.

2012-03-01

The Lie transform formalism is applied to charged particle dynamics in tokamak magnetic topologies, in order to build a Fokker-Planck type operator for Coulomb collisions usable for current drive. This approach makes it possible to reduce the problem to three dimensions (two in velocity space, one in real space) while keeping the wealth of phase-space cross-term coupling effects resulting from conservation of the toroidal canonical momentum (axisymmetry). This kinetic approach makes it possible to describe physical phenomena related to the presence of strong pressure gradients in plasmas of an unspecified form, like the bootstrap current which role will be paramount for the future ITER machine. The choice of coordinates and the method used are particularly adapted to the numerical resolution of the drift kinetic equation making it possible to calculate the particle distributions, which may present a strong variation with respect to the Maxwellian under the effect of an electric field (static or produced by a radio-frequency wave). This work, mainly dedicated to plasma physics of tokamaks, was extended to those of space plasmas with a magnetic dipole configuration. (author)

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)

Ion and electron parameters in the alcator C tokamak scrape-off region

International Nuclear Information System (INIS)

Wan, A.S.H.

1986-05-01

Janus is a bi-directional, multi-functional edge probe used to diagnose the ion and electron parameters in the Alcator C tokamak scrape-off region. Two mirror image sets of diagnostics are aligned to face the electron and ion sides along magnetic field lines. Each set of diagnostics consists of a retarding-field energy analyzer (RFEA), a Langmuir probe, and a calorimeter. The RFEA can alternatively sample both the ion and electron parallel energy distribution functions during a tokamak discharge. From the Langmuir probe, one can infer electron temperature, density, and the plasma floating potential. Simple Langmuir probe theory is found to yield the best agreement between the measured Langmuir probe characteristics and the RFEA-inferred T/sub e/. The calorimeter independently detects the total parallel heat flux incident to an electrically floating plate. The measured sheath transmission coefficient, however, is typically lower than the theoretically predicted value by a factor of approx.3. Together these diagnostics enable detailed, localized edge plasma characterization on Alcator C

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

International Nuclear Information System (INIS)

Reiman, A.

1996-01-01

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

Overvoltage protection for magnetic system during disruption in tokamak

International Nuclear Information System (INIS)

Zhang, Ming; Li, Xiaolong; He, Yang; Zhang, Jun; Chen, Zhongyong; Yu, Kexun

2015-01-01

Highlights: • We investigate the way to limit the plasma disruption overvoltage by using the MOVs. • An overvoltage model of plasma disruption is introduced. • The overvoltage protection scheme has been verified by disruption experiments. • The overvoltage during plasma disruption can be limited to 330 V. - Abstract: During a plasma disruption the magnetic flux in the tokamak changes rapidly, which in most cases will cause high-voltage surges among the magnetic systems and may bring severe damage to the components if there is no overvoltage protection. This paper investigates the way to limit the plasma disruption overvoltage and absorb the energy with the use of metal oxide varistors (MOVs). An overvoltage model of plasma disruption is introduced which can be used for the simulation of plasma disruption and the analysis of the overvoltage. The effectiveness of the overvoltage protection system is validated with disruption experiments. It shows that by optimizing the varistors voltage, the overvoltage during plasma disruption can be limited to an ideal low value. Now the overvoltage protection system has been deployed in J-TEXT tokamak and serves well for daily experiments.

Development of TiC and TiN coated molybdenum limiter system and initial results of the thermal testing in neutral beam heated JFT-2 tokamak

International Nuclear Information System (INIS)

Nakamura, Hiroo; Sengoku, Seio; Maeno, Masaki; Yamamoto, Shin; Seki, Masahiro; Kazawa, Minoru

1982-06-01

This paper describes the limiter drive system for TiC and TiN coated molybdenum limiters and the thermal testing results of the TiC coated limiter in the JFT-2 tokamak using neutral beam injection (0.7 MW). To investigate the influence of TiC coated limiter on plasma behavior and adhesion property under tokamak plasma, a full scale limiter test has been performed in the JFT-2. Reproducible plasma was obtained after the plasma conditioning. Maximum heat flux to the limiter, measured by IR camera, was 1.5 -- 6.5 kW/cm 2 in 25 msec. Cracking, exfoliation and melting on TiC coated limiter were not observed, except for a number of arc tracks. Finally, the permissible heat fluxes of TiC coated molybdenum first wall are discussed. (author)

Vessel eddy current characteristics in SST-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Transport Bifurcation in a Rotating Tokamak Plasma

International Nuclear Information System (INIS)

Highcock, E. G.; Barnes, M.; Schekochihin, A. A.; Parra, F. I.; Roach, C. M.; Cowley, S. C.

2010-01-01

The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

Advantages of iron core in a tokamak

International Nuclear Information System (INIS)

Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

1977-01-01

A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti β (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated

Full Tokamak discharge simulation and kinetic plasma profile control for ITER

International Nuclear Information System (INIS)

Hee Kim, S.

2009-10-01

Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

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

International Nuclear Information System (INIS)

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

1994-09-01

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

Poloidal plasma rotation in the presence of RF waves in tokamaks

International Nuclear Information System (INIS)

Weyssow, B.; Liu, Caigen

2001-01-01

It is well known that one of the consequences of strong RF heating is the deformation of the equilibrium distribution function that induces a change in plasma transport and plasma rotation. The poloidal plasma rotation during RF wave heating in tokamaks is investigated using a moment approach. A set of closed, self-consistent transport and rotation equations is derived and reduced to a single equation for the poloidal particle flux. The formulas are sufficiently general to apply to heating schemes that can be represented by a quasilinear operator. (author)

Progress of the ECH·ECCD experiments. Research progress of the ECH·ECCD experiments in tokamaks and spherical tokamaks

International Nuclear Information System (INIS)

Isayama, Akihiko; Tanaka, Hitoshi

2009-01-01

Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)

Task III: auxillary heating in tokamaks and tandem mirrors. Progress report on fusion plasma theory

International Nuclear Information System (INIS)

Scharer, J.E.

1986-06-01

The research we have accomplished with this grant has focused on ICRF coupling, wave propagation, heating and breakeven studies for tokamaks such as JET. The highlights include fundamental work on a differential equation for wave fields incorporating equilibrium gradients, strong absorption and mode conversion and a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas. We have also formulated and developed a code which solves differential equation for ICRF waveguide coupling in tokamak edge density regions. We are also examining the excitation of ion Bernstein waves from fast magnetosonic waves occurring in density gradients. Our current efforts involve the explanation of current JET ICRF results such as the large electron sawteeth in the core region in terms of hot, non-Maxwellian ICRF theory

Preliminary Design of Alborz Tokamak

Science.gov (United States)

Mardani, M.; Amrollahi, R.; Saramad, S.

2012-04-01

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

Auxiliary radiofrequency heating of tokamaks, Task 3

International Nuclear Information System (INIS)

Scharer, J.E.

1991-07-01

The research performed under this grant during the past three years has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling and heating issues: efficient coupling during the L- to H-mode transition by analysis and computer simulation of ICRF antennas edge plasma profiles; analysis of both dielectric-filled waveguide and coil ICRF antenna coupling to plasma edge profiles; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results; ICRF full-wave field solutions, power conservation and heating analyses; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report. 15 refs

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

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

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

Moving Divertor Plates in a Tokamak

Energy Technology Data Exchange (ETDEWEB)

S.J. Zweben, H. Zhang

2009-02-12

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

Neutronics design for a spherical tokamak fusion-transmutation reactor

International Nuclear Information System (INIS)

Deng Meigen; Feng Kaiming; Yang Bangchao

2002-01-01

Based on studies of the spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. By using the one-dimension transport and burn-up code BISON3.0 to process optimized design, a set of plasma parameters and blanket configuration suitable for the transmutation of MA (Minor Actinides) nuclear waste is selected. Based on the one-dimension calculation, two-dimension calculation has been carried out by using two-dimension neutronics code TWODANT. Combined with the neutron flux given by TWODANT calculation, burn-up calculation has been processed by using the one-dimension radioactivity calculation code FDKR and some useful and reasonable results are obtained

Start-up simulations of the PULSAR pulsed tokamak reactor

International Nuclear Information System (INIS)

Werley, K.A.; Bathke, C.G.

1993-01-01

Start-up conditions are examined for a pulsed tokamak reactor that uses only inductively driven plasma current (and bootstrap current). A zero-dimensional (profile-averaged) model containing plasma power and particle balance equations is used to study several aspects of plasma start-up, including: (1) optimization of the start-up pathway; (2) tradeoffs of auxiliary start-up heating power versus start-up time; (3) volt-second consumption; (4) thermal stability of the operating point; (5) estimates of the diverter heat flux and temperature during the start-up transient; (6) the sensitivity of the available operating space to allowed values of the H confinement factor; and (7) partial-power operations

Magnetohydrodynamic stability of tokamak plasmas with poloidal mode coupling

International Nuclear Information System (INIS)

Shigueoka, H.; Sakanaka, P.H.

1987-01-01

The stability behavior with respect to internal modes is examined for a class of tokamak equilibria with non-circular cross sections. The surfaces of the constant poloidal magnetic flux ψ (R,Z) are obtained numerically by solving the Grad-Shafranov's equation with a specified shape for the outmost plasma surface. The equation of motion for ideal MHD stability is written in a ortogonal coordinate system (ψ, χ, φ). Th e stability analysis is performance numerically in a truncated set of coupled m (poloidal wave number) equations. The calculations involve no approximations, and so all parameters of the equilibrium solution can be arbitrarily varied. (author) [pt

Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

Energy Technology Data Exchange (ETDEWEB)

Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y. [eds.

1996-12-01

This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage.

Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

International Nuclear Information System (INIS)

Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y.

1996-12-01

This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

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

Deuterium retention and surface modification of tungsten macrobrush samples exposed in FTU Tokamak

Science.gov (United States)

Maddaluno, G.; Giacomi, G.; Rufoloni, A.; Verdini, L.

2007-06-01

The effect of discrete structures such as macrobrush or castellated surfaces on power handling and deuterium retention of plasma facing components is to be assessed since such geometrical configurations are needed for increasing the lifetime of the armour to heat-sink joint. Four small macrobrush W and W + 1%La2O3 samples have been exposed in the Frascati Tokamak Upgrade (FTU) scrape-off layer up to the last closed flux surface by means of the Sample Introduction System. FTU is an all metal machine with no carbon source inside vacuum vessel; it exhibits ITER relevant energy and particle fluxes on the plasma facing components. Here, results on morphological surface changes (SEM), chemical composition (EDX) and deuterium retention (TDS) are reported.

Program of thermonuclear reactor structure materials study at Kazakhstan tokamak KTM

International Nuclear Information System (INIS)

Shkolnik, V.S.; Velikhov, E.P.; Cherepnin, Yu. S.; Tikhomirov, L. N.; Tazhibaeva, I.L.; Shestacov, V.P.; Azizov, E.A.; Gostev, A.A.; Buzhinskij, O.A.

2000-01-01

Physical and technical capacities of KTM tokamak are basis of the project. These properties will help to perform a wide spectrum of research on the first wall materials, limiter materials, as well as on materials of divertor plates and mockups of divertor receivers including porous ones with liquid metal cooling within the range of flux loads from 0.1 to 20 MW/m 2 . In research program for the first wall materials the basic attention will be drawn to erosion resistance, recycling, permeability, heat resistance, spraying, possibility of conditioning and recovering their first wall protective properties, material influence on physical processes in hot plasma thread. In the course of limiter material studying basic efforts will be focused on these materials influence on plasma effective charge Z e ff and operation capacity of limiters in a wide spectrum of flux loads

Bi-Maxwellian electron energy distribution function in the vicinity of the last closed flux surface in fusion plasma

Czech Academy of Sciences Publication Activity Database

Popov, T.S.V.K.; Dimitrova, Miglena; Pedrosa, M. A.; López-Bruna, D.; Horáček, Jan; Kovačič, J.; Dejarnac, Renaud; Stöckel, Jan; Aftanas, Milan; Böhm, Petr; Bílková, Petra; Hidalgo, C.; Pánek, Radomír

2015-01-01

Roč. 57, č. 11 (2015), č. článku 115011. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : COMPASS tokamak, parallel power flux density * TJ-II stellarator * bi-Maxwellian EEDF * last closed flux surface * SOL * parallel power flux density Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.404, year: 2015

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

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

The disparate impact of the ion temperature gradient and the density gradient on edge transport and the low-high transition in tokamaks

International Nuclear Information System (INIS)

Kleva, Robert G.; Guzdar, Parvez N.

2009-01-01

Steepening of the ion temperature gradient in nonlinear fluid simulations of the edge region of a tokamak plasma causes a rapid degradation in confinement. As the density gradient steepens, there is a continuous improvement in confinement analogous to the low (L) to high (H) transition observed in tokamaks. In contrast, as the ion temperature gradient steepens, there is a rapid increase in the particle and energy fluxes and no L-H transition. For a given pressure gradient, confinement always improves when more of the pressure gradient arises from the density gradient, and less of the pressure gradient arises from the ion temperature gradient.

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

International Nuclear Information System (INIS)

2000-01-01

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

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

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

Science.gov (United States)

Wróblewski, Dariusz

1997-01-01

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

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

International Nuclear Information System (INIS)

Wroblewski, D.

1997-01-01

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

Status of the COMPASS tokamak and characterization of the first H-mode

Science.gov (United States)

Pánek, R.; Adámek, J.; Aftanas, M.; Bílková, P.; Böhm, P.; Brochard, F.; Cahyna, P.; Cavalier, J.; Dejarnac, R.; Dimitrova, M.; Grover, O.; Harrison, J.; Háček, P.; Havlíček, J.; Havránek, A.; Horáček, J.; Hron, M.; Imríšek, M.; Janky, F.; Kirk, A.; Komm, M.; Kovařík, K.; Krbec, J.; Kripner, L.; Markovič, T.; Mitošinková, K.; Mlynář, J.; Naydenkova, D.; Peterka, M.; Seidl, J.; Stöckel, J.; Štefániková, E.; Tomeš, M.; Urban, J.; Vondráček, P.; Varavin, M.; Varju, J.; Weinzettl, V.; Zajac, J.; the COMPASS Team

2016-01-01

This paper summarizes the status of the COMPASS tokamak, its comprehensive diagnostic equipment and plasma scenarios as a baseline for the future studies. The former COMPASS-D tokamak was in operation at UKAEA Culham, UK in 1992-2002. Later, the device was transferred to the Institute of Plasma Physics of the Academy of Sciences of the Czech Republic (IPP AS CR), where it was installed during 2006-2011. Since 2012 the device has been in a full operation with Type-I and Type-III ELMy H-modes as a base scenario. This enables together with the ITER-like plasma shape and flexible NBI heating system (two injectors enabling co- or balanced injection) to perform ITER relevant studies in different parameter range to the other tokamaks (ASDEX-Upgrade, DIII-D, JET) and to contribute to the ITER scallings. In addition to the description of the device, current status and the main diagnostic equipment, the paper focuses on the characterization of the Ohmic as well as NBI-assisted H-modes. Moreover, Edge Localized Modes (ELMs) are categorized based on their frequency dependence on power density flowing across separatrix. The filamentary structure of ELMs is studied and the parallel heat flux in individual filaments is measured by probes on the outer mid-plane and in the divertor. The measurements are supported by observation of ELM and inter-ELM filaments by an ultra-fast camera.

Optimized tokamak power exhaust with double radiative feedback in ASDEX Upgrade

Science.gov (United States)

Kallenbach, A.; Bernert, M.; Eich, T.; Fuchs, J. C.; Giannone, L.; Herrmann, A.; Schweinzer, J.; Treutterer, W.; the ASDEX Upgrade Team

2012-12-01

A double radiative feedback technique has been developed on the ASDEX Upgrade tokamak for optimization of power exhaust with a standard vertical target divertor. The main chamber radiation is measured in real time by a subset of three foil bolometer channels and controlled by argon injection in the outer midplane. The target heat flux is in addition controlled by nitrogen injection in the divertor private flux region using either a thermoelectric sensor or the scaled divertor radiation obtained by a bolometer channel in the outer divertor. No negative interference of the two radiation controllers has been observed so far. The combination of main chamber and divertor radiative cooling extends the operational space of a standard divertor configuration towards high values of P/R. Pheat/R = 14 MW m-1 has been achieved so far with nitrogen seeding alone as well as with combined N + Ar injection, with the time-averaged divertor peak heat flux below 5 MW m-2. Good plasma performance can be maintained under these conditions, namely H98(y,2) = 1 and βN = 3.

Application of the radiating divertor approach to innovative tokamak divertor concepts

Energy Technology Data Exchange (ETDEWEB)

Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Groebner, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); La Haye, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Leonard, A.W.; Luce, T.C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Maingi, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Moyer, R.A. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Solomon, W.M. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Turco, F. [Columbia University, 2960 Broadway, New York, NY 10027 (United States); Watkins, J.G. [Sandia National Laboratory, PO Box 5800, Albuquerque, NM 87185 (United States)

2015-08-15

We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q{sub ⊥p}) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β{sub N} ≅ 3.0 and H{sub 98(y,2)} ≈ 1.35. It is also demonstrated that q{sub ⊥p} could be reduced ≈50% by extending the parallel connection length (L{sub ||-XPT}) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested.

Probing of the peripheral plasma of a tokamak by diaphragm thermography

International Nuclear Information System (INIS)

Gyunter, K.; Grunov, K.; Evstigneev, S.A.

1985-01-01

The tokamak limiter diverts significant particle and energy fluxes from the plasma and sharply decreases the interaction between the plasma and wall, thus protecting the wall from destruction and the plasma from catastrophic contamination. The temperature of a moveble limiter is measured in the device on the basis of the infrared radiation from its surface. By applying certain models of the layer near the wall this permits to calculate the magnitude of the heat flux directed to the limiter. The power density is, in normal discharges, 2-3 kW/cm 2 ; in discharges with disruptions the power density may reach 10 kW/cm 2 . The measurements show that longitudinal energy flux in the peripheral layer of the plasma filament decreases radially in an exponential manner and can yield the parameter of the decrease. It is found that during discharges involving breakdown a very nonuniform heating of the limiter surface occurs at the initial and final stages of normal discharges, this indicates the existence of a certain structure of the peripheral plasma layer

TRAIL: a tokamak rail gun limiter

International Nuclear Information System (INIS)

Yu, W.S.; Powell, J.R.; Usher, J.L.

1980-01-01

An attractive new limiter concept is investigated. The TRAIL (Tokamak Rail Gun Limiter) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled after cooling, to the injector of an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm 2 can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state of the art (several Km/sec). Accelerators injecting pellets at approx. 1 Km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs

Tokamak Simulation Code modeling of NSTX

International Nuclear Information System (INIS)

Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.

2000-01-01

The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption

Energy confinement and transport of H-mode plasmas in tokamak

International Nuclear Information System (INIS)

Urano, Hajime

2005-02-01

A characteristic feature of the high-confinement (H-mode) regime is the formation of a transport barrier near the plasma edge, where steepening of the density and temperature gradients is observed. The H-mode is expected to be a standard operation mode in a next-step fusion experimental reactor, called ITER-the International Thermonuclear Experimental Reactor. However, energy confinement in the H-mode has been observed to degrade with increasing density. This is a critical constraint for the operation domain in the ITER. Investigation of the main cause of confinement degradation is an urgent issue in the ITER Physics Research and Development Activity. A key element for solving this problem is investigation of the energy confinement and transport properties of H-mode plasmas. However, the influence of the plasma boundary characterized by the transport barrier in H-modes on the energy transport of the plasma core has not been examined sufficiently in tokamak research. The aim of this study is therefore to investigate the energy confinement properties of H-modes in a variety of density, plasma shape, seed impurity concentration, and conductive heat flux in the plasma core using the experimental results obtained in the JT-60U tokamak of Japan Atomic Energy Research Institute. Comparison of the H-mode confinement properties with those of other tokamaks using an international multi-machine database for extrapolation to the next step device was also one of the main subjects in this study. Density dependence of the energy confinement properties has been examined systematically by separating the thermal stored energy into the H-mode pedestal component determined by MHD stability called the Edge Localized Modes (ELMs) and the core component governed by gyro-Bohm-like transport. It has been found that the pedestal pressure imposed by the destabilization of ELM activities led to a reduction in the pedestal temperature with increasing density. The core temperature for each

Research using small tokamaks

International Nuclear Information System (INIS)

1991-01-01

The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

A fluid model for the edge pressure pedestal height and width in tokamaks based on the transport constraint of particle, energy, and momentum balance

Energy Technology Data Exchange (ETDEWEB)

Stacey, W. M., E-mail: weston.stacey@nre.gatech.edu [Georgia Institute of Technology, Atlanta, Georgia 30332 (United States)

2016-06-15

A fluid model for the tokamak edge pressure profile required by the conservation of particles, momentum and energy in the presence of specified heating and fueling sources and electromagnetic and geometric parameters has been developed. Kinetics effects of ion orbit loss are incorporated into the model. The use of this model as a “transport” constraint together with a “Peeling-Ballooning (P-B)” instability constraint to achieve a prediction of edge pressure pedestal heights and widths in future tokamaks is discussed.

Accelerator technology in tokamaks

International Nuclear Information System (INIS)

Kustom, R.L.

1977-01-01

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

Development of hard X-ray spectrometer with high time resolution on the J-TEXT tokamak

Energy Technology Data Exchange (ETDEWEB)

Ma, T.K.; Chen, Z.Y., E-mail: zychen@hust.edu.cn; Huang, D.W.; Tong, R.H.; Yan, W.; Wang, S.Y.; Dai, A.J.; Wang, X.L.

2017-06-01

A hard X-ray (HXR) spectrometer has been developed to study the runaway electrons during the sawtooth activities and during the runaway current plateau phase on the J-TEXT tokamak. The spectrometer system contains four NaI scintillator detectors and a multi-channel analyzer (MCA) with 0.5 ms time resolution. The dedicated peak detection circuit embedded in the MCA provides a pulse height analysis at count rate up to 1.2 million counts per second (Mcps), which is the key to reach the high time resolution. The accuracy and reliability of the system have been verified by comparing with the hardware integrator of HXR flux. The temporal evolution of HXR flux in different energy ranges can be obtained with high time resolution by this dedicated HXR spectrometer. The response of runaway electron transport with different energy during the sawtooth activities can be studied. The energy evolution of runaway electrons during the plateau phase of runaway current can be obtained. - Highlights: • A HXR spectrometer with high time resolution has been developed on J-TEXT tokamak. • The response of REs transport during the sawtooth activities can be investigated. • The energy evolution of REs following the disruptions can be monitored.

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1993-04-01

This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

START: the creation of a spherical tokamak

International Nuclear Information System (INIS)

Sykes, Alan

1992-01-01

The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

Tokamaks (Second Edition)

Energy Technology Data Exchange (ETDEWEB)

Stott, Peter [JET, UK (United Kingdom)

1998-10-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

Tokamaks (Second Edition)

International Nuclear Information System (INIS)

Stott, Peter

1998-01-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

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

International Nuclear Information System (INIS)

Hwang, Y.S.

1993-01-01

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

Geodesic mode instability driven by electron and ion fluxes during neutral beam injection in tokamaks

Czech Academy of Sciences Publication Activity Database

Camilo de Souza, F.; Elfimov, A.; Galvão, R.M.O.; Krbec, Jaroslav; Seidl, Jakub; Stöckel, Jan; Hron, Martin; Havlíček, Josef; Mitošinková, Klára

2017-01-01

Roč. 381, č. 36 (2017), s. 3066-3070 ISSN 0375-9601 R&D Projects: GA ČR(CZ) GA16-25074S; GA ČR(CZ) GA14-35260S; GA MŠk(CZ) 8D15001; GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : Tokamak * Geodesic acoustic modes * Kinetic theory * Instability * Landau damping Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 1.3 Physical sciences Impact factor: 1.772, year: 2016 http://www.sciencedirect.com/science/article/pii/S0375960117306989

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

Magnetic confinement experiment -- 1: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1994-01-01

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

Potential minimum cost of electricity of superconducting coil tokamak power reactors

International Nuclear Information System (INIS)

Reid, R.L.; Peng, Y-K. M.

1989-01-01

The potential minimum cost of electricity (COE) for superconducting tokamak power reactors is estimated by increasing the physics (confinement, beta limit, bootstrap current fraction) and technology [neutral beam energy, toroidal field (TF) coil allowable stresses, divertor heat flux, superconducting coil critical field, critical temperature, and quench temperature rise] constraints far beyond those assumed for ITER until the point of diminishing returns is reached. A version of the TETRA systems code, calibrated with the ITER design and modified for power reactors, is used for this analysis, limiting this study to reactors with the same basic device configuration and costing algorithms as ITER. A minimum COE is reduced from >200 to about 80 mill/kWh when the allowable design constraints are raised to 2 times those of ITER. At 4 times the ITER allowables, a minimum COE of about 60 mill/kWh is obtained. The corresponding tokamak has a major radius of approximately 4 m, a plasma current close to 10 MA, an aspect ratio of 4, a confinement H- factor ≤3, a beta limit of approximately 2 times the first stability regime, a divertor heat flux of about 20 MW/m 2 , a Β max ≤ 18 T, and a TF coil average current density about 3 times that of ITER. The design constraints that bound the minimum COE are the allowable stresses in the TF coil, the neutral beam energy, and the 99% bootstrap current (essentially free current drive). 14 refs., 4 figs., 2 tabs

Self-consistent perturbed equilibrium with neoclassical toroidal torque in tokamaks

International Nuclear Information System (INIS)

Park, Jong-Kyu; Logan, Nikolas C.

2017-01-01

Toroidal torque is one of the most important consequences of non-axisymmetric fields in tokamaks. The well-known neoclassical toroidal viscosity (NTV) is due to the second-order toroidal force from anisotropic pressure tensor in the presence of these asymmetries. This work shows that the first-order toroidal force originating from the same anisotropic pressure tensor, despite having no flux surface average, can significantly modify the local perturbed force balance and thus must be included in perturbed equilibrium self-consistent with NTV. The force operator with an anisotropic pressure tensor is not self-adjoint when the NTV torque is finite and thus is solved directly for each component. This approach yields a modified, non-self-adjoint Euler-Lagrange equation that can be solved using a variety of common drift-kinetic models in generalized tokamak geometry. The resulting energy and torque integral provides a unique way to construct a torque response matrix, which contains all the information of self-consistent NTV torque profiles obtainable by applying non-axisymmetric fields to the plasma. This torque response matrix can then be used to systematically optimize non-axisymmetric field distributions for desired NTV profiles. Published by AIP Publishing.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Microwave Tokamak Experiment: Overview and status

International Nuclear Information System (INIS)

1990-05-01

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

Runaway relativistic electron scattering on the plazma oscillations in tokamak

International Nuclear Information System (INIS)

Krasovitskij, V.B.; Razdorski, V.G.

1980-01-01

The dynamics of fast electrons in a tolamak plasma with the presence of the constant external electric field have been inveatigated. It is shown that the occurrence of the relativistic electrons ''tail'' of the distribution function is followed by an intensive plasma oscillation swinging under conditions of the anomalous Doppler effect and their large angle scattering in the momentum space. A part of scattered electrons is captured by tokamak inhomogeneous magnetic field and causes the occurrence of a new low frequency alfven instability under conditions of magnetic drift resonance followed by quasilinear diffusion of relativistic electrons along the small radius of the torus. The flux of runaway electrons scattered on plasma oscillations has been found. A nonlinear diffusion equation has been derived for the flux of captured electrons. The equation defines the carrying out of fast particles from the plasma filament center to its periphery depending on the external magnetic field and plasma parameters

Multichordal visible/near-UV spectroscopy on the DIII-D tokamak

International Nuclear Information System (INIS)

Seraydarian, R.P.; Burrell, K.H.; Groebner, R.J.

1988-01-01

A pair of visible/near-UV spectrometers with eight viewing chords apiece have been installed on the DIII-D tokamak. Each system views a neutral heating beam and can acquire up to 250 complete spectra from each chord with 5--20-ms time resolution. Each viewing chord covers 60 A with 0.27-A spectral resolution, and the chords span about (2)/(3) of the plasma's full width. By viewing Doppler-broadened spectral lines from charge exchange recombination (CER) reactions between beam neutrals and plasma ions, ion temperatures up to 4 keV have been measured, and the bulk Doppler shift of these same lines has yielded plasma rotation velocities up to 200 km/s. The constancy of temperature on a magnetic flux surface and the rigid rotor model of a flux surface have been confirmed. These instruments have also been used to measure the neutral beam deposition profile, and preliminary experimental results agree with theoretical calculations of the beam deposition profile

Multichordal visible/near uv spectroscopy in the DIII-D tokamak

International Nuclear Information System (INIS)

Seraydarian, R.P.; Burrell, K.H.; Groebner, R.J.

1988-02-01

A pair of visible/near uv spectrometers with eight viewing chords apiece have been installed on the DIII-D tokamak. Each system views a neutral heating beam, and can acquire up to 250 complete spectra from each chord with 5-20 msec time resolution. Each viewing chord covers 60 A with 0.27 A spectral resolution, and the chords span about 2/3 of the plasma's full width. By viewing Doppler broadened spectral lines from charge exchange recombination (CER) reactions between beam neutrals and plasma ions, ion temperatures up to keV have been meassured, and the bulk Doppler shift of these same lines has yielded plasma rotation velocities up to 200 km/sec. The constancy of temperature on a magnetic flux surface and the rigid rotor model of a flux surface have been confirmed. These instruments have also been used to measure the neutral beam deposition profile, and preliminary experimental results agree with theoretical calculations of the beam deposition profile. 5 refs., 6 figs

Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas

Czech Academy of Sciences Publication Activity Database

Horáček, Jan; Pitts, R.A.; Adámek, Jiří; Arnoux, G.; Bak, J.-G.; Brezinsek, S.; Dimitrova, Miglena; Goldston, R.J.; Gunn, J. P.; Havlíček, Josef; Hong, S.-H.; Janky, Filip; LaBombard, B.; Marsen, S.; Maddaluno, G.; Nie, L.; Pericoli, V.; Popov, Tsv.; Pánek, Radomír; Rudakov, D.; Seidl, Jakub; Seo, D.S.; Shimada, M.; Silva, C.; Stangeby, P.C.; Viola, B.; Vondráček, Petr; Wang, H.; Xu, G.S.; Xu, Y.

2016-01-01

Roč. 58, č. 7 (2016), č. článku 074005. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2011021 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tokamak * ITER * SOL decay length * SOL width * scaling Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/0741-3335/58/7/074005

Neutral particle dynamics in the Alcator C-Mod tokamak

Energy Technology Data Exchange (ETDEWEB)

Niemczewski, Artur P. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

1995-08-01

This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism.

Neutral particle dynamics in the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Niemczewski, A.P.

1995-08-01

This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism. 146 refs., 82 figs., 14 tabs

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-01-01

This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

Magnetic ''islandography'' in tokamaks

International Nuclear Information System (INIS)

Callen, J.D.; Waddell, B.V.; Hicks, H.R.

1978-09-01

Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N.; Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N.; Lebedev, V.M.; Litunovstkii, N.V.; Mazul, I.

2007-01-01

Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm 3 . The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities ∼ 10 20 m -3 . This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material exposed to prolonged

Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

Energy Technology Data Exchange (ETDEWEB)

Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N. [A.F. IOFFE Physico-technical Institute, Russian Academy of Sciences, St Petersburg (Russian Federation); Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N. [St. Petersburg State Univ., Research Institute of Physics (Russian Federation); Lebedev, V.M. [B.P. Konstantinov Nuclear Physics Institute, Russian Academy of Science, Gatchina (Russian Federation); Litunovstkii, N.V. [D.V. Efremov Institute of Electrophysical Apparatus, St.Petersburg (Russian Federation); Mazul, I. [Development of Plasma Facing Materials and Components Laboratory, EFREMOV INSTITUTE, St Petersbourg (Russian Federation)

2007-07-01

Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm{sup 3}. The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities {approx} 10{sup 20} m{sup -3}. This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material

MHD simulations of DC helicity injection for current drive in tokamaks

International Nuclear Information System (INIS)

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

1994-12-01

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

Characterisation of detached plasmas on the MAST tokamak

Energy Technology Data Exchange (ETDEWEB)

Harrison, J.R., E-mail: james.harrison@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Lisgo, S.W. [ITER Organization, Route de Vinon-sur-Verdon, St.Paul-lez-Durance, Cedex (France); Gibson, K.J. [Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Dowling, J. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)

2011-08-01

Divertor detachment is an attractive operating regime for the next generation of tokamak devices, as it offers a means of mitigating the steady-state heat flux to plasma facing components. In order to clarify the dominant physical mechanisms that govern detachment, high quality data from several diagnostics are required to constrain theoretical models. To that end, high spatial ({approx}3 mm) and temporal (5 kHz) resolution measurements have been made of the intensity of deuterium Balmer and carbon emission lines during the onset and evolution of detachment of the lower inner strike point in MAST L-mode discharges. Furthermore, spatially-resolved measurements of the shapes and intensities of high-n Balmer lines have been recorded to infer plasma conditions during the detached phase.

The Mercier Criterion in Reversed Shear Tokamak Plasmas

International Nuclear Information System (INIS)

Kessel, C.; Chance, M.S.; Jardin, S.C.

1999-01-01

A recent numerical study has found that, contrary to conventional theoretical and experimental expectations, reversed shear plasmas are unstable primarily because the term proportional to the shear in the Mercier criterion is destabilizing. In the present study, the role of the magnetic shear, both local and global, is examined for various tokamak configurations with monotonic and non-monotonic safety factor profiles. The enhancement of the local shear due to the outward shift of the magnetic axis suggests that the latter are less susceptible to interchanges. Furthermore, by regrouping the terms in the criterion, the V'' term when differentiated instead with respect to the toroidal flux, is shown to absorb the dominant shear term. No Mercier instability is found for similar profiles as in the previous study

Two-dimensional transport of tokamak plasmas

International Nuclear Information System (INIS)

Hirshman, S.P.; Jardin, S.C.

1979-01-01

A reduced set of two-fluid transport equations is obtained from the conservation equations describing the time evolution of the differential particle number, entropy, and magnetic fluxes in an axisymmetric toroidal plasma with nested magnetic surfaces. Expanding in the small ratio of perpendicular to parallel mobilities and thermal conductivities yields as solubility constraints one-dimensional equations for the surface-averaged thermodynamic variables and magnetic fluxes. Since Ohm's law E +u x B =R', where R' accounts for any nonideal effects, only determines the particle flow relative to the diffusing magnetic surfaces, it is necessary to solve a single two-dimensional generalized differential equation, (partial/partialt) delpsi. (delp - J x B) =0, to find the absolute velocity of a magnetic surface enclosing a fixed toroidal flux. This equation is linear but nonstandard in that it involves flux surface averages of the unknown velocity. Specification of R' and the cross-field ion and electron heat fluxes provides a closed system of equations. A time-dependent coordinate transformation is used to describe the diffusion of plasma quantities through magnetic surfaces of changing shape

Anomalous Convection Reversal due to Turbulence Transition in Tokamak Plasmas

International Nuclear Information System (INIS)

Sun Tian-Tian; Chen Shao-Yong; Huang Jie; Mou Mao-Lin; Tang Chang-Jian; Wang Zhan-Hui; Peng Xiao-Dong

2015-01-01

A critical physical model, based on the ion temperature gradient (ITG) mode and the trapped electron mode (TEM), trying to explain the spatio-temporal dynamics of anomalous particle convection reversal (i.e., the particle convective flux reverses from inward to outward), is developed numerically. The dependence of density peaking and profile shape on the particle convection is studied. Only the inward pinch could lead to the increase of the density peaking. The validation of the critical model is also analyzed. A comparison of the estimates calculated by the model and the experimental results from the Tore Supra tokamak shows that they are qualitatively both consistent. (paper)

Conceptual design study of the moderate size superconducting spherical tokamak power plant

Science.gov (United States)

Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

2015-06-01

A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

A need for non-tokamak approaches to magnetic fusion energy

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

Reiman, A.

1995-05-01

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

Energy transport requirements for tokamak reactors in the second ballooning stability regime

International Nuclear Information System (INIS)

Potok, R.E.; Bromberg, L.; Cohn, D.R.

1986-01-01

The authors present an analysis of ignition confinement constraints on a tokamak reactor operating in the second regime of ballooning stability. This regime is characterized by flat plasma pressure profiles, with a sharp pressure gradient near a conducting first wall at the plasma edge. The energy confinement time is determined by transport processes across the pressure gradient region. The authors have found that the required transport needed for ignition in the edge region is very close to the value predicted by neoclassical ion conductivity scaling. Only by carefully tailoring the conductivity scaling across the flux coordinate were the authors able to match both the kink stability and ignition requirements. With optimistic assumptions, R/sub o/ ≅ 7 m appears to be the minimum major radius for an economical tokamak reactor in the second ballooning stability regime. This paper presents a base design case at R/sub o/ = 7 m, and shows how the reactor design varies with changes in major radius, ion transport scaling, and electron transport scaling

Density peaking in the JFT-2M tokamak plasma with counter neutral beam injection

International Nuclear Information System (INIS)

Ida, K.; Itoh, S.; Itoh, K.

1991-05-01

A significant particle pinch and reduction of the effective thermal diffusivity are observed after switching the neutral beam direction from co- to counter- injection in the JFT-2M tokamak. A time delay in the occurrence of density peaking to that of plasma rotation is found. This shows that the particle pinch is related to the profile of the electric field as determined by the plasma rotation profile. The measured particle flux shows qualitative agreement with the theoretically-predicted inward pinch. (author)

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1985-01-01

The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1986-01-01

Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

The role of the spherical tokamak in clarifying tokamak physics

International Nuclear Information System (INIS)

Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

1999-01-01

The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

Reactor similarity for plasma–material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

International Nuclear Information System (INIS)

Whyte, D.G.; Olynyk, G.M.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Hartwig, Z.S.; Mumgaard, R.T.; Podpaly, Y.A.

2012-01-01

Highlights: ► Discussion of similarity scalings for reduced-size tokamaks. ► Proposal of a new set of scaling laws for divertor similarity. ► Discussion of how the new scaling provides fidelity to a reactor. ► The new scaling is used as the basis for the Vulcan conceptual design. - Abstract: Dimensionless parameter scaling techniques are a powerful tool in the study of complex physical systems, especially in tokamak fusion experiments where the cost of full-size devices is high. It is proposed that dimensionless similarity be used to study in a small-scale device the coupled issues of the scrape-off layer (SOL) plasma, plasma–material interactions (PMI), and the plasma-facing material (PFM) response expected in a tokamak fusion reactor. Complete similarity is not possible in a reduced-size device. In addition, “hard” technological limits on the achievable magnetic field and peak heat flux, as well as the necessity to produce non-inductive scenarios, must be taken into account. A practical approach is advocated, in which the most important dimensionless parameters are matched to a reactor in the reduced-size device, while relaxing those parameters which are far from a threshold in behavior. “Hard” technological limits are avoided, so that the reduced-size device is technologically feasible. A criticism on these grounds is offered of the “P/R” model, in which the ratio of power crossing the last closed flux surface (LCFS), P, to the device major radius, R, is held constant. A new set of scaling rules, referred to as the “P/S” scaling (where S is the LCFS area) or the “PMI” scaling, is proposed: (i) non-inductive, steady-state operation; (ii) P is scaled with R 2 so that LCFS areal power flux P/S is constant; (iii) magnetic field B constant; (iv) geometry (elongation, safety factor q * , etc.) constant; (v) volume-averaged core density scaled as n≈n ¯ e ∼R −2/7 ; and (vi) ambient wall material temperature T W,0 constant. It is

Fast modeling of flux trapping cascaded explosively driven magnetic flux compression generators.

Science.gov (United States)

Wang, Yuwei; Zhang, Jiande; Chen, Dongqun; Cao, Shengguang; Li, Da; Liu, Chebo

2013-01-01

To predict the performance of flux trapping cascaded flux compression generators, a calculation model based on an equivalent circuit is investigated. The system circuit is analyzed according to its operation characteristics in different steps. Flux conservation coefficients are added to the driving terms of circuit differential equations to account for intrinsic flux losses. To calculate the currents in the circuit by solving the circuit equations, a simple zero-dimensional model is used to calculate the time-varying inductance and dc resistance of the generator. Then a fast computer code is programmed based on this calculation model. As an example, a two-staged flux trapping generator is simulated by using this computer code. Good agreements are achieved by comparing the simulation results with the measurements. Furthermore, it is obvious that this fast calculation model can be easily applied to predict performances of other flux trapping cascaded flux compression generators with complex structures such as conical stator or conical armature sections and so on for design purpose.

Advanced tokamak burning plasma experiment

International Nuclear Information System (INIS)

Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

2001-01-01

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

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)

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)

Pellet-plasma interaction: Local disturbances caused by pellet ablation in tokamaks

International Nuclear Information System (INIS)

Lengyel, L.L.

1989-01-01

The local disturbance amplitudes caused by ablating pellets in tokamaks are computed in the framework of a magnetohydrodynamic model supplemented by the neutral gas plasma shielding ablation model. The model computes, for a given number of pellet particles locally deposited, the time histories of the ablatant cloud parameters, such as cloud radius, cloud length, electron density, temperature and cloud beta, at a succession of magnetic flux surfaces. The cloud radius thus determined may be fed back into the ablation model, thus adjusting the effect of the shielding cloud on the ablation rate. The model is applied to typical plasma parameter ranges of existing and planned tokamaks. The results show that the ablating pellets may cause massive local disturbances in tokamaks, depending upon the number of particles locally deposited. The peaks of these disturbances are of a spike nature, lasting only a few microseconds (Alfven time-scale). The characteristic decay time of the 'quasi-steady' disturbance values that characterize the after-spike period is of the order of several milliseconds (hydrodynamic time-scale). The peak electron density values may be as high as 10 23 to 10 25 m -3 , with the associated beta peaks exceeding unity. The 'quasi-steady' values of the electron density and the ablatant beta may be of the order of 10 22 to 10 24 m -3 and unity, respectively. Furthermore, the results show the strong dependence of the ablation rate on the dynamic characteristics of the ablatant cloud surrounding the pellet. (author). 25 refs, 6 figs, 2 tabs

Resistive requirements for the vacuum wall of a tokamak fusion reactor

International Nuclear Information System (INIS)

Brooks, J.N.; Ehat, D.; Harkness, S.D.; Norem, J.; Stevens, H.; Turner, L.

1978-01-01

Most conceptual designs of tokamak power reactors have incorporated a ceramic insulator in the vacuum wall to make the wall electrically non-conducting. Such a material will have to be highly resistant to radiation damage at doses up to at least 10 MW-yr/m 2 while being compatible with a coolant and a first wall whose dimensions change due to thermal cycling and radiation damage. Thus there is considerable incentive to assess the consequences of eliminating the flux breaker from the design and having a conducting boundary instead. In this initial study the question of having a finite wall resistance has been examined in terms of its major implications on both the normal and abnormal operation of a tokamak reactor. This study has been conducted within the framework of the ANL-EPR-77 design although the results should provide some guidance for future reactors as well. The EPR design referred to is a 5 m major radius tokamak with an aspect ratio of 3.5, and with an equilibrium plasma current of 7.3 MA. The vacuum chamber is designed to accommodate a non-circular plasma with a height to width ratio of up to 1.65. The basic vacuum wall design is shown in Fig. 1. It is located about 0.4 M from the plasma boundary and has an irregular polygon shape made of sixteen sections, one per TF coil interval. Variations of this design having a range of resistance values have been used in the analysis

Research into controlled fusion in tokamaks

International Nuclear Information System (INIS)

Zacek, F.

1992-01-01

During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

Application studies of spherical tokamak plasma merging

International Nuclear Information System (INIS)

Ono, Yasushi; Inomoto, Michiaki

2012-01-01

The experiment of plasma merging and heating has long history in compact torus studies since Wells. The study of spherical tokamak (ST), starting from TS-3 plasma merging experiment of Tokyo University in the late 1980s, is followed by START of Culham laboratory in the 1900s, TS-4 and UTST of Tokyo University and MAST of Culham laboratory in the 2000s, and last year by VEST of Soul University. ST has the following advantages: 1) plasma heating by magnetic reconnection at a MW-GW level, 2) rapid start-up of high beta plasma, 3) current drive/flux multiplication and distribution control of ST plasma, 4) fueling and helium-ash exhaust. In the present article, we emphasize that magnetic reconnection and plasma merging phenomena are important in ST plasma study as well as in plasma physics. (author)

Development of fast video recording of plasma interaction with a lithium limiter on T-11M tokamak

Energy Technology Data Exchange (ETDEWEB)

Lazarev, V.B., E-mail: v_lazarev@triniti.ru [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Dzhurik, A.S.; Shcherbak, A.N. [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Belov, A.M. [NRC “Kurchatov Institute”, Moscow (Russian Federation)

2016-11-15

Highlights: • The paper presents the results of the study of tokamak plasma interaction with lithium capillary-porous system limiters and PFC by high-speed color camera. • Registration of emission near the target in SOL in neutral lithium light and e-folding length for neutral Lithium measurements. • Registration of effect of MHD instabilities on CPS Lithium limiter. • A sequence of frames shows evolution of lithium bubble on the surface of lithium limiter. • View of filament structure near the plasma edge in ohmic mode. - Abstract: A new high-speed color camera with interference filters was installed for fast video recording of plasma-surface interaction with a Lithium limiter on the base of capillary-porous system (CPS) in T-11M tokamak vessel. The paper presents the results of the study of tokamak plasma interaction (frame exposure time up to 4 μs) with CPS Lithium limiter in a stable stationary phase, unstable regimes with internal disruption and results of processing of the image of the light emission around the probe, i.e. e-folding length for neutral Lithium penetration and e-folding length for Lithium ion flux in SOL region.

The density limit in Tokamaks

International Nuclear Information System (INIS)

Alladio, F.

1985-01-01

A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

Development of Operation Scenario for Spherical Tokamak at SNU

International Nuclear Information System (INIS)

Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

2009-01-01

Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

Vlasov modelling of parallel transport in a tokamak scrape-off layer

International Nuclear Information System (INIS)

Manfredi, G; Hirstoaga, S; Devaux, S

2011-01-01

A one-dimensional Vlasov-Poisson model is used to describe the parallel transport in a tokamak scrape-off layer. Thanks to a recently developed 'asymptotic-preserving' numerical scheme, it is possible to lift numerical constraints on the time step and grid spacing, which are no longer limited by, respectively, the electron plasma period and Debye length. The Vlasov approach provides a good velocity-space resolution even in regions of low density. The model is applied to the study of parallel transport during edge-localized modes, with particular emphasis on the particles and energy fluxes on the divertor plates. The numerical results are compared with analytical estimates based on a free-streaming model, with good general agreement. An interesting feature is the observation of an early electron energy flux, due to suprathermal electrons escaping the ions' attraction. In contrast, the long-time evolution is essentially quasi-neutral and dominated by the ion dynamics.

Vlasov modelling of parallel transport in a tokamak scrape-off layer

Energy Technology Data Exchange (ETDEWEB)

Manfredi, G [Institut de Physique et Chimie des Materiaux, CNRS and Universite de Strasbourg, BP 43, F-67034 Strasbourg (France); Hirstoaga, S [INRIA Nancy Grand-Est and Institut de Recherche en Mathematiques Avancees, 7 rue Rene Descartes, F-67084 Strasbourg (France); Devaux, S, E-mail: Giovanni.Manfredi@ipcms.u-strasbg.f, E-mail: hirstoaga@math.unistra.f, E-mail: Stephane.Devaux@ccfe.ac.u [JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

2011-01-15

A one-dimensional Vlasov-Poisson model is used to describe the parallel transport in a tokamak scrape-off layer. Thanks to a recently developed 'asymptotic-preserving' numerical scheme, it is possible to lift numerical constraints on the time step and grid spacing, which are no longer limited by, respectively, the electron plasma period and Debye length. The Vlasov approach provides a good velocity-space resolution even in regions of low density. The model is applied to the study of parallel transport during edge-localized modes, with particular emphasis on the particles and energy fluxes on the divertor plates. The numerical results are compared with analytical estimates based on a free-streaming model, with good general agreement. An interesting feature is the observation of an early electron energy flux, due to suprathermal electrons escaping the ions' attraction. In contrast, the long-time evolution is essentially quasi-neutral and dominated by the ion dynamics.

Module of lithium divertor for KTM tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

Fast Waves Mode Conversion and Energy Deposition in Simulated, Pre-Heated, Neoclassical, Tight Aspect Ratio Tokamak Plasmas

International Nuclear Information System (INIS)

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

1999-01-01

Some basic aspects of wave-plasma interaction of interest for tight aspect ratio spherical tokamaks are investigated theoretically. The following scenario is considered: A. Fast magnetosonic waves are launched by an external antenna into a simulated spherical Tokamak plasma; these waves are converted to Alfven waves at points (layer) satisfying the Alfven resonance condition. B. The simulated spherical tokamaks-plasma has a circular cross-section and toroidicity effects are simulated by Grad-Shafranov type, radially dependent axial magnetic field and its shear. (J. Actual equilibrium profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. D. The study is based on the numerical solution of the full e.m. wave equation which includes a quite general resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. Two kinds of results will be presented: I. Proofs validating the computational algorithm used and including convergence and energy conservation. II. Exact quantitative results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited p over . The dependence of the results on the launched wave characteristics (wave numbers, frequency and intensity) as well as on those of the equilibrium plasma (equilibrium current, neoclassical resistivity and electron inertia) will be discussed

Economic considerations of commercial tokamak options

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m 2 , which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m 2 , will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e)

Statistical theory of L-H transition in tokamaks

International Nuclear Information System (INIS)

Itoh, S.-I.; Itoh, K.; Toda, S.

2003-01-01

A statistical model for the bifurcation of the radial electric field E r is developed in view of describing L-H transitions of tokamak plasmas. Noise in micro fluctuations is shown to lead to random changes of E r , if a deterministic approach allows for more than one solution. The probability density function for and the ensemble average of E r are obtained. The L-to-H and the H-to-L transition probabilities are calculated, and the effective phase limit is derived. Due to the suppression of turbulence by shear in E r , the limit deviates from Maxwell's rule. The ensemble average of heat flux as well as that of E r do not show a hysteresis in contrast to the deterministic model. Experimental condition for observing the hysteresis is also addressed. (author)

Theory-based transport simulation of tokamaks: density scaling

International Nuclear Information System (INIS)

Ghanem, E.S.; Kinsey, J.; Singer, C.; Bateman, G.

1992-01-01

There has been a sizeable amount of work in the past few years using theoretically based flux-surface-average transport models to simulate various types of experimental tokamak data. Here we report two such studies, concentrating on the response of the plasma to variation of the line averaged electron density. The first study reported here uses a transport model described by Ghanem et al. to examine the response of global energy confinement time in ohmically heated discharges. The second study reported here uses a closely related and more recent transport model described by Bateman to examine the response of temperature profiles to changes in line-average density in neutral-beam-heated discharges. Work on developing a common theoretical model for these and other scaling experiments is in progress. (author) 5 refs., 2 figs

Large aspect ratio tokamak study

International Nuclear Information System (INIS)

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

1979-01-01

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

Plasma boundary phenomena in tokamaks

International Nuclear Information System (INIS)

Stangeby, P.C.

1989-06-01

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

Nuclear fusion research at Tokamak Energy Ltd

International Nuclear Information System (INIS)

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

2017-01-01

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

Experimental study on plasma parameters in the mushroom limiter shadow in the T-10 tokamak

International Nuclear Information System (INIS)

Alferov, A.A.; Vershkov, V.A.; Grashin, S.A.; Chankin, A.V.

1988-01-01

Plasma parameters in the shadow of mashroom limiter installed in the lower tokamak outlet are studied. Investigation into asymmetry of plasma fluxes to the ion and electrone limiter sides leads to a consumption concerning two meachanisms of its occurrance-toroidal plasma rotation and prevailing plasma departure to the wall through the external torus encirclement. Asymmetry of plasma drift potentials near the limiter observed during the experiment leads to current drift through the limiter close to Spitzer j s one. It is shown that with the increase of mean plasma density the plasma density in the limiter channels grows and its temperature is decreased so the charged particle losses for the limiter are weakly dependent on the mean density which is connected with plasma confinement degradation under the density reduction. A complete flux of charged particles to the limiter is comparable to their flux from plasma filament. Plasma flux into the channels is close to ambipolar one and the power fluxes to neutralization plates are of the order of 10 j s Te/e. Neutral gas pressure dependence in the volume under the limiter on the plasma fluxes to channels is nonlinear, the maximum pressure achieves 3x10 -2 T

Plasma potential and electron temperature evaluated by ball-pen and Langmuir probes in the COMPASS tokamak.

Czech Academy of Sciences Publication Activity Database

Dimitrova, Miglena; Popov, Tsv.K.; Adámek, Jiří; Kovačič, J.; Ivanova, P.; Hasan, E.; López-Bruna, D.; Seidl, Jakub; Vondráček, Petr; Dejarnac, Renaud; Stöckel, Jan; Imríšek, Martin; Pánek, Radomír

2017-01-01

Roč. 59, č. 12 (2017), č. článku 125001. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : Plasma potential * electron temperature * bi-Maxwellian EEDF * ball-pen probe * Langmuir probe * COMPASS tokamak * last closed flux surface Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016

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

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

Mercier criterion for high-β tokamaks

International Nuclear Information System (INIS)

Galvao, R.M.O.

1984-01-01

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

Plasma-neutral gas interaction in a tokamak divertor: effects of hydrogen molecules and plasma recombination

International Nuclear Information System (INIS)

Krasheninnikov, S.I.; Pigarov, A.Yu.; Soboleva, T.K.; Sigmar, D.J.

1997-01-01

We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 x 10 -10 cm 3 /s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a 'gas box' divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (orig.)

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

International Nuclear Information System (INIS)

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

1988-05-01

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

A transmission/escape probabilities model for neutral particle transport in the outer regions of a diverted tokamak

International Nuclear Information System (INIS)

Stacey, W.M.

1992-12-01

A new computational model for neutral particle transport in the outer regions of a diverted tokamak plasma chamber is presented. The model is based on the calculation of transmission and escape probabilities using first-flight integral transport theory and the balancing of fluxes across the surfaces bounding the various regions. The geometrical complexity of the problem is included in precomputed probabilities which depend only on the mean free path of the region

Tokamak Physics Experiment (TPX) design

International Nuclear Information System (INIS)

Schmidt, J.A.

1995-01-01

TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

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

Application of SNAM to the nuclear analysis of EAST Tokamak

International Nuclear Information System (INIS)

Hu, H.; Chen, M.; Zeng, Q.; Wu, Y.

2007-01-01

EAST (Experimental Advanced Superconducting Tokamak) is the first non-round cross section complete superconducting fusion experimental tokamak device built at the Institute of Plasma Physics, Chinese Academy of Sciences. Since 2.45MeV neutrons from D-D fusion reaction and 14.1MeV neutrons from D-T fusion reaction can both be generated during the DD plasma discharge, the distribution of neutron flux and nuclear heat in the device has an important effect on the nuclear and safety analysis. For radiation transport calculations, the main calculation tool is Monte Carlo transport code (MCNP) which has been used to give specific nuclear responses in complex geometries. However, the discrete ordinate transport code (SN code) is more effective to calculate the distribution of neutron flux and nuclear heat in the whole device. It is a time-consuming and error prone task to prepare the neutronics model for SN code in manual way. And because of the self-limitation of SN method, most of SN codes only support some specified or regular geometries, such as cylindrical, Cartesian and tetrahedral geometry, etc. In practice, most of models are composed of irregular solids and can not be supported by SN code. A more efficient solution is to shift the geometric modeling into a computer aided design (CAD) system and use an interface for SN code to convert CAD model into the input file automatically. SNAM (SN Code Automatic Modeling) is an integrated interface code between CAD system and SN code. The CAD model can be automatically converted into the input file of SN code with SNAM. On the contrary, SNAM can convert already existed input file of SN code into CAD model, which can be used to check, analyze, modify and reuse the model for the user. In this contribution, the process of converting EAST CAD model to the input file of SN code with SNAM is described. The distribution of neutron flux and nuclear heat in the whole device are calculated using the input file, and the results of the

Summary of the 1982 small tokamak users meeting

International Nuclear Information System (INIS)

Sprott, J.C.

1982-11-01

On November 1, 1982, the sixth in a series of approximately annual meetings of the users of small tokamaks was held in conjunction with the APS Division of Plasma Physics meeting at New Orleans. The meeting lasted three hours, with 34 people attending. The interest was on strengthening the ties between the small tokamaks and the large tokamaks. Accordingly, the latest meeting was dedicated to this theme, and in contrast to previous meetings, a few representatives from the large tokamaks were invited to attend and make presentations. Summaries of the various talks are included

Magnetic confinement by Tokamak: physical aspects

International Nuclear Information System (INIS)

Tachon, J.

1980-01-01

After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr

Resistive instabilities in tokamaks

International Nuclear Information System (INIS)

Rutherford, P.H.

1985-10-01

Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed

Tokamak research in the Soviet Union

International Nuclear Information System (INIS)

Strelkov, V.S.

1981-01-01

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

Neutral beam in ALVAND IIC tokamak

International Nuclear Information System (INIS)

Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

1992-01-01

Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

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

International Nuclear Information System (INIS)

Mikhailovskii, A.B.

1998-01-01

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

Modular pump limiter systems for large tokamaks

International Nuclear Information System (INIS)

Uckan, T.; Klepper, C.C.; Mioduszewski, P.K.; McGrath, R.T.

1987-09-01

Long-pulse (>10-s) operation of large tokamaks with high-power (>10-MW) heating and extensive external fueling will require correspondingly efficient particle exhaust for density control. A pump limiter can provide the needed exhaust capability by removing a small percentage of the particles, which would otherwise be recycled. Single pump limiter modules have been operated successfully on ISX-B, PDX, TEXTOR, and PLT. An axisymmetric pump limiter is now being installed and will be studied in TEXTOR. A third type of pump limiter is a system that consists of several modules and exhibits performance different from that of a single module. To take advantage of the flexibility of a modular pump limiter system in a high-power, long-pulse device, the power load must be distributed among a number of modules. Because each added module changes the performance of all the others, a set of design criteria must be defined for the overall limiter system. The design parameters for the modules are then determined from the system requirements for particle and power removal. Design criteria and parameters are presented, and the impact on module design of the state of the art in engineering technology is discussed. The relationship between modules are considered from the standpoint of flux coverage and shadowing effects. The results are applied to the Tore Supra tokamak. A preliminary conceptual design for the Tore Supra pump limiter system is discussed, and the design parameters of the limiter modules are presented. 21 refs., 12 figs

Physics design and experimental study of tokamak divertor

International Nuclear Information System (INIS)

Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

2007-06-01

The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

Diagnostics for the Rijnhuizen Tokamak Project

NARCIS (Netherlands)

Donne, A. J. H.

1994-01-01

The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel

Plasma position control in TCABR Tokamak

International Nuclear Information System (INIS)

Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

1998-01-01

The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

Global gyrokinetic simulation of tokamak transport

International Nuclear Information System (INIS)

Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T.

1998-10-01

A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or η i (η i ≡ ∂ ell nT i /∂ ell n n i ) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling

Slowing down tail enhanced, neoclassical and classical alpha particle fluxes in tokamak reactors

International Nuclear Information System (INIS)

Catto, P.J.; Tessarotto, M.

1988-01-01

The classical and neoclassical particle and energy fluxes associated with a slowing down tail, alpha particle distribution function are evaluated for arbitrary aspect ratio ε -1 , cross section, and poloidal magnetic field. The retention of both electron and ion drag and pitch angle scattering by the background ions results in a large diffusive neoclassical heat flux in the plasma core. This flux remains substantial at larger radii only if the characteristic speed associated with pitch angle scattering, v/sub b/, is close enough to the alpha birth speed v 0 so that ε(v 0 /v/sub b/) 3 remains less than some order unity critical value which is not determined by the methods herein. The enhanced neoclassical losses would only have a serious impact on ignition if the critical value of ε(v 0 /v/sub b/) 3 is found to be somewhat larger than unity

Presheath profiles in simulated tokamak edge plasmas

International Nuclear Information System (INIS)

LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

1988-04-01

The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

Robust Sliding Mode Control for Tokamaks

Directory of Open Access Journals (Sweden)

I. Garrido

2012-01-01

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

Three novel tokamak plasma regimes in TFTR

International Nuclear Information System (INIS)

Furth, H.P.

1985-10-01

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

Kinetic and collision process effects on magnetic structures in pre-disruption phase of tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Farshi, Esmaeil [Kyushu Univ., Advanced Energy Engineering Sciences, Kasuga, Fukuoka (Japan); Goudarzi, Shervin [AEOI, Plasma Physics Department, Tehran (Iran); Amrollahi, Reza [K-N Toosi Univ. of Technology, Tehran (Iran); Sato, Kohnosuke [Kyushu Univ., Research Institute for Applied Mechanics, Kasuga, Fukuoka (Japan)

2001-07-01

Oscillations of the parallel and perpendicular neutral fluxes that are observed during pre-disruption stage in recent experiments, show possibility of a structure in pre-disruption phase of tokamak plasmas. This structure oscillates simultaneously with the m=2 mode until the damping of this mode. The perpendicular component of this structure is greater than the parallel one. From other side, there are a good correlation between MHD activity and behavior of charge exchange neutrals, and an enough good correlation between time behavior of charge exchange flux with high energy and OV line radiation in pre-disruption phase. These may witness possibility of a mechanism of losses-excitation of inner transition with help of heavy particles in pre-disruption phase. This mechanism plays an important role in magnetic structures in pre-disruption phase. (author)

Kinetic and collision process effects on magnetic structures in pre-disruption phase of tokamak plasmas

International Nuclear Information System (INIS)

Farshi, Esmaeil; Goudarzi, Shervin; Amrollahi, Reza; Sato, Kohnosuke

2001-01-01

Oscillations of the parallel and perpendicular neutral fluxes that are observed during pre-disruption stage in recent experiments, show possibility of a structure in pre-disruption phase of tokamak plasmas. This structure oscillates simultaneously with the m=2 mode until the damping of this mode. The perpendicular component of this structure is greater than the parallel one. From other side, there are a good correlation between MHD activity and behavior of charge exchange neutrals, and an enough good correlation between time behavior of charge exchange flux with high energy and OV line radiation in pre-disruption phase. These may witness possibility of a mechanism of losses-excitation of inner transition with help of heavy particles in pre-disruption phase. This mechanism plays an important role in magnetic structures in pre-disruption phase. (author)

Internal disruption in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

Internal disruptions in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

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

International Nuclear Information System (INIS)

Kim, Yil Bong.

1988-02-01

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

Flux consumption optimization and the achievement of 1 MA discharges on NSTX

International Nuclear Information System (INIS)

Menard, J.; LeBlanc, B.; Sabbagh, S.A.

2001-01-01

The spherical tokamak (ST), because of its slender central column, has very limited volt-second capability relative to a standard aspect ratio tokamak of similar plasma cross-section. Recent experiments on the National Spherical Torus Experiment (NSTX) have begun to quantify and optimize the ohmic current drive efficiency in a MA-class ST device. Sustainable ramp-rates in excess of 5MA/sec during the current rise phase have been achieved on NSTX, while faster ramps generate significant MHD activity. Discharges with I P exceeding 1MA have been achieved in NSTX with nominal parameters: aspect ratio A=1.3-1.4, elongation κ=2-2.2, triangularity δ=0.4, internal inductance l i =0.6, and Ejima coefficient C E =0.35. Flux consumption efficiency results, performance improvements associated with first boronization, and comparisons to neoclassical resistivity are described. (author)

Flux-driven turbulence GDB simulations of the IWL Alcator C-Mod L-mode edge compared with experiment

Science.gov (United States)

Francisquez, Manaure; Zhu, Ben; Rogers, Barrett

2017-10-01

Prior to predicting confinement regime transitions in tokamaks one may need an accurate description of L-mode profiles and turbulence properties. These features determine the heat-flux width upon which wall integrity depends, a topic of major interest for research aid to ITER. To this end our work uses the GDB model to simulate the Alcator C-Mod edge and contributes support for its use in studying critical edge phenomena in current and future tokamaks. We carried out 3D electromagnetic flux-driven two-fluid turbulence simulations of inner wall limited (IWL) C-Mod shots spanning closed and open flux surfaces. These simulations are compared with gas puff imaging (GPI) and mirror Langmuir probe (MLP) data, examining global features and statistical properties of turbulent dynamics. GDB reproduces important qualitative aspects of the C-Mod edge regarding global density and temperature profiles, within reasonable margins, and though the turbulence statistics of the simulated turbulence follow similar quantitative trends questions remain about the code's difficulty in exactly predicting quantities like the autocorrelation time A proposed breakpoint in the near SOL pressure and the posited separation between drift and ballooning dynamics it represents are examined This work was supported by DOE-SC-0010508. This research used resources of the National Energy Research Scientific Computing Center (NERSC).

Low-energy hydrogen flux measurements at the TORTUR tokamak with negative ion conversion

International Nuclear Information System (INIS)

Toledo, Wiebo van.

1990-01-01

The interaction of a tokamak plasma with the vessel wall is one of the most important subjects in thermonuclear research. The information about this interaction is not complete without direct detection of the outward stream of low-energy, down to a few electronvolts, neutral hydrogen or deuterium atoms. The detection of these atoms is the subject of this thesis. An appropriate method to analyse the atoms which are emitted from the edge plasma is to use a time-of-flight analyser. This kind of apparatus selects particles according to their velocities with-out distinguishing between different masses. If these analysers use the Daly-method the lowest measurable energy of the hydrogen atoms is approximately 25 electronvolts. To increase the detection efficiency a new detection method was developed. This new method uses the conversion of hydrogen atoms into H- ions on a cesiated tungsten surface. By this conversion the lowest measurable energy is decreased down to 5 electron-volt. (author). 93 refs.; 44 figs.; 7 tabs

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.

Statistical theory of L-H transition in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Itoh, S.-I. [Kyushu Univ., Research Institute for Applied Mechanics, Kasuga, Fukuoka (Japan); Itoh, K.; Toda, S. [National Inst. for Fusion Science, Toki, Gifu (Japan)

2003-01-01

A statistical model for the bifurcation of the radial electric field E{sub r} is developed in view of describing L-H transitions of tokamak plasmas. Noise in micro fluctuations is shown to lead to random changes of E{sub r}, if a deterministic approach allows for more than one solution. The probability density function for and the ensemble average of E{sub r} are obtained. The L-to-H and the H-to-L transition probabilities are calculated, and the effective phase limit is derived. Due to the suppression of turbulence by shear in E{sub r}, the limit deviates from Maxwell's rule. The ensemble average of heat flux as well as that of E{sub r} do not show a hysteresis in contrast to the deterministic model. Experimental condition for observing the hysteresis is also addressed. (author)

Contour analysis of steady state tokamak reactor performance

International Nuclear Information System (INIS)

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

1990-01-01

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

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.

1984-05-01

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

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.; California Univ., Los Angeles

1984-01-01

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

Estimation of Zeff in Novillo Tokamak

International Nuclear Information System (INIS)

Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

1996-01-01

We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

Energy Technology Data Exchange (ETDEWEB)

McGrath, R.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Yamashina, T. [ed.] [Hokkadio Univ. (Japan)

1994-04-01

This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition.

US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

International Nuclear Information System (INIS)

McGrath, R.T.; Yamashina, T.

1994-04-01

This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition

Neutral molecules in tokamak edge plasma - role of vibrationally excited hydrogen molecules

International Nuclear Information System (INIS)

Cadez, I.; Cercek, M.; Pelicon, P.; Razpet, A.

2003-01-01

The role of neutral molecules in edge plasma is discussed with special emphasis on the vibrationally excited hydrogen. Neutral molecules are formed mostly by surface processes on the walls and then released to the edge plasma where they take part in volumetric reactions with other particles. Typically these molecules are formed in excited states and data are needed for their reactions on the wall and in the volume. Processes in edge plasma determine particle and energy flux what is especially critical issue in tokamak divertor region. Various cross sections and reaction rates are needed for modelling edge plasma and its interaction with walls. (author)

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

Confinement and diffusion in tokamaks

International Nuclear Information System (INIS)

McWilliams, R.

1988-01-01

The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

Numerical studies of edge localized instabilities in tokamaks

International Nuclear Information System (INIS)

Wilson, H.R.; Snyder, P.B.; Huysmans, G.T.A.; Miller, R.L.

2002-01-01

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code

Axisymmetric instability in a noncircular tokamak: experiment and theory

International Nuclear Information System (INIS)

Lipschultz, B.; Prager, S.C.; Todd, A.M.M.; Delucia, J.

1979-09-01

The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes -- the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria. Experimentally, the square is vertically stable and both dee's unstable to a vertical nonrigid axisymmetric shift. The central magnetic axis displacement grows exponentially with a growth time approximately 10 3 poloidal Alfven times plasma time. Proper initial positioning of the plasma on the midplane allows passive feedback to nonlinearly restore vertical motion to a small stable oscillation. Experimental poloidal flux plots are produced directly from internal magnetic probe measurements

Propagation of Blob in boundary of HT-7 tokamak

International Nuclear Information System (INIS)

Yan Ning; Zhang Wei; Chang Jiafeng; Ming Tingfeng; Ding Siye

2011-01-01

Intermittent characteristics of turbulence induced by coherent structures (Blob) are manifested clearly on the Langmuir probe signal of HT-7 tokamak. With conditional analysis, asymmetric characteristics of the intermittent bursts are demonstrated. The parameter of plasma inside the Blob is larger than the background plasma parameter. Due to the radial propagation of the coherent structures, the particle density and temperature profiles in the scrape-off layer (SOL) are non-exponential and flat away from the last closed flux surface (LCFS). Around the LCFS, large burst fluctuations are responsible for about 50% of the total transport. These experimental finds may imply that the coherent structures are distorted by the developed shear flow in the shear layer. In SOL region, the coherent structures propagate in the direction of ion diamagnetic drift. (authors)

An overview on plasma disruption mitigation and avoidance in tokamak

International Nuclear Information System (INIS)

He Kaihui; Pan Chuanhong; Feng Kaiming

2002-01-01

Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

TRANSPORT BY INTERMITTENCY IN THE BOUNDARY OF THE DIII-D TOKAMAK

International Nuclear Information System (INIS)

BOEDO, JA; RUDAKOV, DL; MOYER, RA; MCKEE, GR; COLCHIN, RJ; SCHAFFER, MJ; STANGEBY, PG; WEST, WP; ALLEN, SL; EVANS, TE; FONCK, RJ; HOLLMANN, EM; KRASHENINNIKOV, S; LEONARD, AW; NEVINS, W; MAHDAVI, MA; PORTER, GD; TYNAN, GR; WHYTE, DG; XU, X

2002-01-01

A271 TRANSPORT BY INTERMITTENCY IN THE BOUNDARY OF THE DIII-D TOKAMAK. Intermittent plasma objectives (IPOs) featuring higher pressure than the surrounding plasma, are responsible for ∼ 50% of the E x B T radial transport in the scrape off layer (SOL) of the DIII-D tokamak in L- and H-mode discharges. Conditional averaging reveals that the IPOs are positively charged and feature internal poloidal electric fields of up to 4000 V/m. The IPOs move radially with E x B T /B 2 velocities of ∼ 2600 m/s near the last closed flux surface (LCFS), and ∼ 330 m/s near the wall. The IPOs slow down as they shrink in radial size from 4 cm at the LCFS to 0.5 cm near the wall. The skewness (i.e. asymmetry of fluctuations from the average) of probe and beam emission spectroscopy (BES) data indicate IPO formation at or near the LCFS and the existence of positive and negative IPOs which move in opposite directions. The particle content of the IPOs at the LCFS is linearly dependent on the local density and decays over ∼ 3 cm into the SOL while their temperature decays much faster (∼ 1 cm)

Probe measurements of hydrogen fluxes during discharge cleaning in JFT-2M

International Nuclear Information System (INIS)

Matsuzaki, Y.

1989-01-01

Thermal desorption spectroscopy (TDS) has been applied during discharge cleaning in the JFT-2M tokamak to measure hydrogen fluxes. The TDS carbon sample, thickness 0.13 mm, was heated to 1000 0 C by direct current and the temperature distribution of the sample surface measured by infrared thermography. The probe was exposed to three types of plasma: Taylor-type discharge cleaning (TDC), ECR discharge cleaning (ECR-DC), and glow discharge cleaning (GDC). The TDS spectra show peak desorption at around 800 0 C. The hydrogen flux, obtained by integration of the TDS spectrum, decreases exponentially in the radial direction with decay length 7.4 cm and 5.8 cm in TDC and ECR-DC, respectively. The relation between hydrogen fluxes and water vapour production was investigated. In TDC, the amount of water vapour depends more strongly on the electron temperature of the plasma than on the hydrogen flux. In ECR-DC, the production of water vapour increases approximately linearly with the hydrogen-flux. In GDC, hydrogen fluxes were measured by TDS but no water vapour could be detected in the residual gases during the discharge. (orig.)

Note: Measurement of the runaway electrons in the J-TEXT tokamak

International Nuclear Information System (INIS)

Chen, Z. Y.; Zhang, Y.; Zhang, X. Q.; Luo, Y. H.; Jin, W.; Li, J. C.; Chen, Z. P.; Wang, Z. J.; Yang, Z. J.; Zhuang, G.

2012-01-01

The runaway electrons have been measured by hard x-ray detectors and soft x-ray array in the J-TEXT tokamak. The hard x-ray radiations in the energy ranges of 0.5-5 MeV are measured by two NaI detectors. The flux of lost runaway electrons can be obtained routinely. The soft x-ray array diagnostics are used to monitor the runaway beam generated in disruptions since the soft x-ray is dominated by the interaction between runaway electrons and metallic impurities inside the plasma. With the aid of soft x-ray array, runaway electron beam has been detected directly during the formation of runaway current plateau following the disruptions.

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.

Small angle slot divertor concept for long pulse advanced tokamaks

Science.gov (United States)

Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

2017-04-01

SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

Tearing instability in a tokamak with a noncircular cross section

Energy Technology Data Exchange (ETDEWEB)

Arsenin, V. V., E-mail: arsenin-vv@nrcki.ru; Skovoroda, A. A., E-mail: skovoroda-aa@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

2016-12-15

Using a straight-column model to describe tokamak plasma with a noncircular cross section, it is shown how to (i) find the boundary of tearing instability from the condition of existence of a magnetohydrodynamic plasma equilibrium different from that of a straight cylinder by solving a two-dimensional linear boundary-value problem with a second-order equation with respect to the flux coordinate and (ii) find the spatial structure of the tearing mode and the corresponding effective Δ' when there is only one resonance magnetic surface in the plasma for a given axial wavenumber by solving some kind of a boundary-value problem for the perturbation. The proposed approach is illustrated by numerical calculations for the case of an elliptical cross section as an example.

Tokamak COMPASS

Czech Academy of Sciences Publication Activity Database

Řípa, Milan; Křenek, Petr

2011-01-01

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

Analysis of equilibrium and topology of tokamak plasmas

International Nuclear Information System (INIS)

Milligen, B.P. van.

1991-01-01

In a tokamak, the plasma is confined by means of a magnetic field. There exists an equilibrium between outward forces due to the pressure gradient in plasma and inward forces due to the interaction between currents flowing inside the plasma and the magnetic field. The equilibrium magnetic field is characterized by helical field lines that lie on nested toroidal surfaces of constant flux. The equilibrium yields values for global and local plasma parameters (e.g. plasma position, total current, local pressure). Thus, precise knowledge of the equilibrium is essential for plasma control, for the understanding of many phenomena occurring in the plasma (in particular departures from the ideal equilibrium involving current filamentation on the flux surfaces that lead to the formation of islands, i.e. nested helical flux surfaces), and for the interpretation of many different types of measurements (e.g. the translation of line integrated electron density measurements made by laser beams probing the plasma into a local electron density on a flux surface). The problem of determining the equilibrium magnetic field from external magnetic field measurements has been studied extensively in literature. The problem is 'ill-posed', which means that the solution is unstable to small changes in the measurement data, and the solution has to be constrained in order to stabilize it. Various techniques for handling this problem have been suggested in literature. Usually ad-hoc restrictions are imposed on the equilibrium solution in order to stabilize it. More equilibrium solvers are not able to handle very dissimilar measurement data which means information on the equilibrium is lost. The generally do not allow a straightforward error estimate of the obtained results to be made, and they require large amounts of computing time. This problems are addressed in this thesis. (author). 104 refs.; 42 figs.; 6 tabs

An enhanced tokamak startup model

Science.gov (United States)

Goswami, Rajiv; Artaud, Jean-François

2017-01-01

The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

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

Energy Technology Data Exchange (ETDEWEB)

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

1978-06-01

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

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

International Nuclear Information System (INIS)

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

1978-06-01

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

A constant heat flux plasma limiter for TEXTOR

International Nuclear Information System (INIS)

Mioduszewski, P.

1980-10-01

In future large tokamak machines heat removal from the plasma is going to play an important role. In TEXTOR the total plasma power is expected to be in the range of 0.5-2.5 MW. Typical fractions of about 50% of this power have to be removed from the plasma by limiters. The power flux from the limiter scrape-off layer to the limiter surface decays rapidly with distance into the scrape-off layer resulting in a highly space-dependent heat load on the limiter. Therefore, limiters are shaped in a way to smooth of the heat load, and the ideal limiter shape should produce a constant heat flux over the whole limiter surface. The ideally shaped limiter offers a better chance to handle the high heat loads with the preferred materials like stainless steel (or inconel 625 as in the case of TEXTOR). (orig./GG)

Magnetic flux pumping in 3D nonlinear magnetohydrodynamic simulations

Science.gov (United States)

Krebs, I.; Jardin, S. C.; Günter, S.; Lackner, K.; Hoelzl, M.; Strumberger, E.; Ferraro, N.

2017-10-01

A self-regulating magnetic flux pumping mechanism in tokamaks that maintains the core safety factor at q ≈1 , thus preventing sawteeth, is analyzed in nonlinear 3D magnetohydrodynamic simulations using the M3D-C1 code. In these simulations, the most important mechanism responsible for the flux pumping is that a saturated (m =1 ,n =1 ) quasi-interchange instability generates an effective negative loop voltage in the plasma center via a dynamo effect. It is shown that sawtoothing is prevented in the simulations if β is sufficiently high to provide the necessary drive for the (m =1 ,n =1 ) instability that generates the dynamo loop voltage. The necessary amount of dynamo loop voltage is determined by the tendency of the current density profile to centrally peak which, in our simulations, is controlled by the peakedness of the applied heat source profile.

Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

2003-01-01

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

Enhancement of confinement in tokamaks

International Nuclear Information System (INIS)

Furth, H.P.

1986-01-01

The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

2003-12-01

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

Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

2003-01-01

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

Bibliography of fusion product physics in tokamaks

International Nuclear Information System (INIS)

Hively, L.M.; Sigmar, D.J.

1989-09-01

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

Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

International Nuclear Information System (INIS)

Sheffield, J.

1985-01-01

During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

Deposition of deuterium and metals on divertor tiles in the DIII-D tokamak

International Nuclear Information System (INIS)