Edge plasma physical investigations of tokamak plasmas in CRIP
International Nuclear Information System (INIS)
Bakos, J.; Ignacz, P.; Koltai, L.; Paszti, F.; Petravich, G.; Szigeti, J.; Zoletnik, S.
1988-01-01
The results of the measurements performed in the field of thermonuclear high temperature plasma physics in CRIP (Hungary) are summarized. In the field of the edge plasma physics solid probes were used to test the external zone of plasma edges, and atom beams and balls were used to investigate both the external and internal zones. The plasma density distribution was measured by laser blow-off technics, using Na atoms, which are evaporated by laser pulses. The excitation of Na atom ball by tokamak plasma gives information on the status of the plasma edge. The toroidal asymmetry of particle transport in tokamak plasma was measured by erosion probes. The evaporated and transported impurities were collected on an other part of the plasma edge and were analyzed by SIMS and Rutherford backscattering. The interactions in plasma near the limiter were investigated by a special limiter with implemented probes. Recycling and charge exchange processes were measured. Disruption phenomena of tokamak plasma were analyzed and a special kind of disruptions, 'soft disruptions' and the related preliminary perturbations were discovered. (D.Gy.) 10 figs
Edge Plasma Physics and Relevant Diagnostics on the CASTOR tokamak
Czech Academy of Sciences Publication Activity Database
Stöckel, Jan; Devynck, P.; Gunn, J.; Martines, E.; Bonhomme, G.; Van Oost, G.; Hron, Martin; Ďuran, Ivan; Pánek, Radomír; Stejskal, Pavel; Adámek, Jiří
2004-01-01
Roč. 3, - (2004), s. 1-6 ISSN 1433-5581. [First Cairo Conference on Plasma Physics & Applications. Cairo, 11.10.2003-15.10.2003] R&D Projects: GA ČR GA202/03/0786; GA ČR GP202/03/P062 Keywords : tokamak * edge plasma * probe diagnostics * biasing * turbulence * polarization Subject RIV: BL - Plasma and Gas Discharge Physics
Energy Technology Data Exchange (ETDEWEB)
Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Kocman, J.; Grover, O. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Krbec, J.; Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, CZ-182 21 Prague (Czech Republic)
2015-10-15
Graphical abstract: * Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes.* Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform.* More than 20% plasma life prolongation with plasma position control in feedback mode. - Highlights: • Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes. • Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform. • More than 20% plasma life prolongation with plasma position control in feedback mode. - Abstract: The GOLEM tokamak at the Czech Technical University has been established as an educational tokamak device for domestic and foreign students. Remote participation in the scope of several laboratory practices, plasma physics schools and workshops has been successfully performed from abroad. A new enhancement allowing understandable remote control of vertical plasma position in two modes (i) predefined and (ii) feedback control is presented. It allows to drive the current in the stabilization coils in any time-dependent scenario, which can include as a parameter the actual plasma position measured by magnetic diagnostics. Arbitrary movement of the plasma column in a vertical direction, stabilization of the plasma column in the center of the tokamak vessel as well as prolongation/shortening of plasma life according to the remotely defined request are demonstrated.
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)
International Nuclear Information System (INIS)
Haines, M.G.
1984-01-01
The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)
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
International Nuclear Information System (INIS)
1979-01-01
This report contains the papers delivered at the AEB - Natal University summer school on plasma physics held in Durban during January 1979. The following topics were discussed: Tokamak devices; MHD stability; trapped particles in tori; Tokamak results and experiments; operating regime of the AEB Tokamak; Tokamak equilibrium; high beta Tokamak equilibria; ideal Tokamak stability; resistive MHD instabilities; Tokamak diagnostics; Tokamak control and data acquisition; feedback control of Tokamaks; heating and refuelling; neutral beam injection; radio frequency heating; nonlinear drift wave induced plasma transport; toroidal plasma boundary layers; microinstabilities and injected beams and quasilinear theory of the ion acoustic instability
Real-time control of Tokamak plasmas: from control of physics to physics-based control
International Nuclear Information System (INIS)
Felici, F. A. A.
2011-11-01
Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a digital realtime control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This experimental platform has been used to develop and test new control strategies for three plasma physics instabilities: sawtooth, edge localized mode (ELM) and neoclassical tearing mode (NTM). The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface (q: safety factor). A sawtooth pacing controller was developed which is able to control the time of appearance of the next sawtooth crash. Each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes have been used to study the relationship between sawteeth and NTMs. Postcrash MHD activity can provide the ‘seed’ island for an NTM, which then grows under its neoclassical bootstrap drive. The seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have
Divertor plasma physics experiments on the DIII-D tokamak
International Nuclear Information System (INIS)
Mahdavi, M.A.; Allen, S.L.; Evans, T.E.
1996-10-01
In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model
On the physics of the pressure and temperature gradients in the edge of tokamak plasmas
Stacey, Weston M.
2018-04-01
An extended plasma fluid theory including atomic physics, radiation, electromagnetic and themodynamic forces, external sources of particles, momentum and energy, and kinetic ion orbit loss is employed to derive theoretical expressions that display the role of the various factors involved in the determination of the pressure and temperature gradients in the edge of tokamak plasmas. Calculations for current experiments are presented to illustrate the magnitudes of various effects including strong radiative and atomic physics edge cooling effects and strong reduction in ion particle and energy fluxes due to ion orbit loss in the plasma edge. An important new insight is the strong relation between rotation and the edge pressure gradient.
Spectroscopy and atomic physics of highly ionized Cr, Fe, and Ni for tokamak plasmas
Feldman, U.; Doschek, G. A.; Cheng, C.-C.; Bhatia, A. K.
1980-01-01
The paper considers the spectroscopy and atomic physics for some highly ionized Cr, Fe, and Ni ions produced in tokamak plasmas. Forbidden and intersystem wavelengths for Cr and Ni ions are extrapolated and interpolated using the known wavelengths for Fe lines identified in solar-flare plasmas. Tables of transition probabilities for the B I, C I, N I, O I, and F I isoelectronic sequences are presented, and collision strengths and transition probabilities for Cr, Fe, and Ni ions of the Be I sequence are given. Similarities of tokamak and solar spectra are discussed, and it is shown how the atomic data presented may be used to determine ion abundances and electron densities in low-density plasmas.
International Nuclear Information System (INIS)
1993-12-01
The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support
International Nuclear Information System (INIS)
Hooper, E.B.; Allen, S.L.; Casper, T.A.; Thomassen, K.I.
1989-01-01
This note describes the diagnostics and auxiliary systems on the Microwave Tokamak Experiment (MTX) for confinement, transport, and other plasma physics studies. It is intended as a reference on the installed and planned hardware on the machine for those who need more familiarity with this equipment. Combined with the tokamak itself, these systems define the opportunities and capabilities for experiments in the MTX facility. We also illustrate how these instruments and equipment are to be used in carrying out the MTX Operations Plan. Near term goals for MTX are focussed on the absorption and heating by the microwave beam from the FEL, but the Plan also includes using the facility to study fundamental phenomena in the plasma, to control MHD activity, and to drive current noninductively
Czech Academy of Sciences Publication Activity Database
Svoboda, V.; Kocman, J.; Grover, O.; Krbec, Jaroslav; Stöckel, Jan
96-97, October (2015), s. 974-979 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] Institutional support: RVO:61389021 Keywords : tokamak technology * remote participation * plasma stabilization Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://dx.doi.org/10.1016/j.fusengdes.2015.06.044
The National Spherical Tokamak Experiment at the Princeton Plasma Physics Laboratory
International Nuclear Information System (INIS)
1995-12-01
The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1108, evaluating the environmental effects of the proposed construction and operation of the National Spherical Tokamak Experiment (NSTX) within the existing C-Stellarator (CS) Building at the Princeton Plasma Physics Laboratory, Princeton, New Jersey. The purpose of the NSTX is to investigate the physics of spherically shaped plasmas as an alternative path to conventional tokamaks for development of fusion energy. Fusion energy has the potential to help compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Construction of the NSTX in the CS Building would require the dismantling and removal of the existing unused Princeton Large Torus (PLT) device, part of which would be reused to construct the NSTX. Based on the analyses in the EA, the DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 U.S.C. 4,321 et seq. The preparation of an Environmental Impact Statement is not required. Thus, the DOE is issuing a FONSI pursuant to the Council on Environmental Quality regulations implementing NEPA (40 CFR Parts 1500--1508) and the DOE NEPA implementing regulations (10 CFR Part 1021)
Energy Technology Data Exchange (ETDEWEB)
NONE
1994-05-27
If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).
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
On the HL-1M tokamak plasma confinement time
International Nuclear Information System (INIS)
Qin Yunwen
2001-01-01
Emphasizing that the tokamak plasma confinement time is the plasma particle or thermal energy loss characteristic time, the relevant physical concept and HL-1M tokamak experimental data analyses are reviewed
Plasma diagnostics on large tokamaks
International Nuclear Information System (INIS)
Orlinskij, D.V.; Magyar, G.
1988-01-01
The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs
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)
Energy Technology Data Exchange (ETDEWEB)
Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica
1997-12-31
Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)
Energy Technology Data Exchange (ETDEWEB)
White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.
1989-01-01
The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.
Physics conditions for robust control of tearing modes in a rotating tokamak plasma
Lazzaro, E.; Borgogno, D.; Brunetti, D.; Comisso, L.; Fevrier, O.; Grasso, D.; Lutjens, H.; Maget, P.; Nowak, S.; Sauter, O.; Sozzi, C.; the EUROfusion MST1 Team
2018-01-01
The disruptive collapse of the current sustained equilibrium of a tokamak is perhaps the single most serious obstacle on the path toward controlled thermonuclear fusion. The current disruption is generally too fast to be identified early enough and tamed efficiently, and may be associated with a variety of initial perturbing events. However, a common feature of all disruptive events is that they proceed through the onset of magnetohydrodynamic instabilities and field reconnection processes developing magnetic islands, which eventually destroy the magnetic configuration. Therefore the avoidance and control of magnetic reconnection instabilities is of foremost importance and great attention is focused on the promising stabilization techniques based on localized rf power absorption and current drive. Here a short review is proposed of the key aspects of high power rf control schemes (specifically electron cyclotron heating and current drive) for tearing modes, considering also some effects of plasma rotation. From first principles physics considerations, new conditions are presented and discussed to achieve control of the tearing perturbations by means of high power ({P}{{EC}}≥slant {P}{{ohm}}) in regimes where strong nonlinear instabilities may be driven, such as secondary island structures, which can blur the detection and limit the control of the instabilities. Here we consider recent work that has motivated the search for the improvement of some traditional control strategies, namely the feedback schemes based on strict phase tracking of the propagating magnetic islands.
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
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
Tokamak plasma interaction with limiters
International Nuclear Information System (INIS)
Pitcher, C.S.
1987-11-01
The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D 2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict
International Nuclear Information System (INIS)
Hassanein, A.; Konkashbaev, I.
1999-01-01
The structure of a collisionless scrape-off-layer (SOL) plasma in tokamak reactors is being studied to define the electron distribution function and the corresponding sheath potential between the divertor plate and the edge plasma. The collisionless model is shown to be valid during the thermal phase of a plasma disruption, as well as during the newly desired low-recycling normal phase of operation with low-density, high-temperature, edge plasma conditions. An analytical solution is developed by solving the Fokker-Planck equation for electron distribution and balance in the SOL. The solution is in good agreement with numerical studies using Monte-Carlo methods. The analytical solutions provide an insight to the role of different physical and geometrical processes in a collisionless SOL during disruptions and during the enhanced phase of normal operation over a wide range of parameters
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)
Electron thermal transport in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Konings, J A
1994-11-30
The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).
Reux, Cedric
2017-10-01
Runaway electrons are created during disruptions of tokamak plasmas. They can be accelerated in the form of a multi-MA beam at energies up to several 10's of MeV. Prevention or suppression of runaway electrons during disruptions will be essential to ensure a reliable operation of future tokamaks such as ITER. Recent experiments showed that the suppression of an already accelerated beam with massive gas injection was unsuccessful at JET, conversely to smaller tokamaks. This was attributed to a dense, cold background plasma (up to several 1020 m-3 accompanying the runaway beam. The present contribution reports on the latest experimental results obtained at JET showing that some mitigation efficiency can be restored by changing the features of the background plasma. The density, temperature, position of the plasma and the energy of runaways were characterized using a combined analysis of interferometry, soft X-rays, bolometry, magnetics and hard X-rays. It showed that lower density background plasmas were obtained using smaller amounts of gas to trigger the disruption, leading to an improved penetration of the mitigation gas. Based on the observations, a physical model of the creation of the background plasma and its subsequent evolution is proposed. The plasma characteristics during later stages of the disruption are indeed dependent on the way it was initially created. The sustainment of the plasma during the runaway beam phase is then addressed by making a power balance between ohmic heating, power transfer from runaway electrons, radiation and atomic processes. Finally, a model of the interaction of the plasma with the mitigation gas is proposed to explain why massive gas injection of runaway beams works only in specific situations. This aims at pointing out which parameters bear the most importance if this mitigation scheme is to be used on larger devices like ITER. Acknowledgement: This work has been carried out within the framework of the EUROfusion Consortium
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
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
ECRH Studies on Tokamak Plasmas.
1980-10-10
r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately
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.)
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)
Princeton Plasma Physics Laboratory
Energy Technology Data Exchange (ETDEWEB)
1990-01-01
This report discusses the following topics: principal parameters achieved in experimental devices fiscal year 1990; tokamak fusion test reactor; compact ignition tokamak; Princeton beta experiment- modification; current drive experiment-upgrade; international collaboration; x-ray laser studies; spacecraft glow experiment; plasma processing: deposition and etching of thin films; theoretical studies; tokamak modeling; international thermonuclear experimental reactor; engineering department; project planning and safety office; quality assurance and reliability; technology transfer; administrative operations; PPPL patent invention disclosures for fiscal year 1990; graduate education; plasma physics; graduate education: plasma science and technology; science education program; and Princeton Plasma Physics Laboratory reports fiscal year 1990.
Princeton Plasma Physics Laboratory
International Nuclear Information System (INIS)
1990-01-01
This report discusses the following topics: principal parameters achieved in experimental devices fiscal year 1990; tokamak fusion test reactor; compact ignition tokamak; Princeton beta experiment- modification; current drive experiment-upgrade; international collaboration; x-ray laser studies; spacecraft glow experiment; plasma processing: deposition and etching of thin films; theoretical studies; tokamak modeling; international thermonuclear experimental reactor; engineering department; project planning and safety office; quality assurance and reliability; technology transfer; administrative operations; PPPL patent invention disclosures for fiscal year 1990; graduate education; plasma physics; graduate education: plasma science and technology; science education program; and Princeton Plasma Physics Laboratory reports fiscal year 1990
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
Plasma internal inductance dynamics in a tokamak
International Nuclear Information System (INIS)
Romero, J.A.
2010-01-01
A lumped parameter model for tokamak plasma current and inductance time evolution as a function of plasma resistance, non-inductive current drive sources and boundary voltage or poloidal field coil current drive is presented. The model includes a novel formulation leading to exact equations for internal inductance and plasma current dynamics. Having in mind its application in a tokamak inductive control system, the model is expressed in state space form, the preferred choice for the design of control systems using modern control systems theory. The choice of system states allows many interesting physical quantities such as plasma current, inductance, magnetic energy, and resistive and inductive fluxes be made available as output equations. The model is derived from energy conservation theorem, and flux balance theorems, together with a first order approximation for flux diffusion dynamics. The validity of this approximation has been checked using experimental data from JET showing an excellent agreement.
Princeton Plasma Physics Laboratory:
Energy Technology Data Exchange (ETDEWEB)
Phillips, C.A. (ed.)
1986-01-01
This paper discusses progress on experiments at the Princeton Plasma Physics Laboratory. The projects and areas discussed are: Principal Parameters Achieved in Experimental Devices, Tokamak Fusion Test Reactor, Princeton Large Torus, Princeton Beta Experiment, S-1 Spheromak, Current-Drive Experiment, X-ray Laser Studies, Theoretical Division, Tokamak Modeling, Spacecraft Glow Experiment, Compact Ignition Tokamak, Engineering Department, Project Planning and Safety Office, Quality Assurance and Reliability, and Administrative Operations.
Princeton Plasma Physics Laboratory:
International Nuclear Information System (INIS)
Phillips, C.A.
1986-01-01
This paper discusses progress on experiments at the Princeton Plasma Physics Laboratory. The projects and areas discussed are: Principal Parameters Achieved in Experimental Devices, Tokamak Fusion Test Reactor, Princeton Large Torus, Princeton Beta Experiment, S-1 Spheromak, Current-Drive Experiment, X-ray Laser Studies, Theoretical Division, Tokamak Modeling, Spacecraft Glow Experiment, Compact Ignition Tokamak, Engineering Department, Project Planning and Safety Office, Quality Assurance and Reliability, and Administrative Operations
Relaxed states of tokamak plasmas
International Nuclear Information System (INIS)
Kucinski, M.Y.; Okano, V.
1993-01-01
The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)
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)
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)
Tokamak plasma boundary layer model
International Nuclear Information System (INIS)
Volkov, T.F.; Kirillov, V.D.
1983-01-01
A model has been developed for the limiter layer and for the boundary region of the plasma column in a tokamak to facilitate analytic calculations of the thickness of the limiter layers, the profiles and boundary values of the temperature and the density under various conditions, and the difference between the electron and ion temperatures. This model can also be used to analyze the recycling of neutrals, the energy and particle losses to the wall and the limiter, and other characteristics
Plasma edge physics in the TEXTOR tokamak with poloidal and toroidal limiters
International Nuclear Information System (INIS)
Samm, U.; Bogen, P.; Hartwig, H.; Hintz, E.; Hoethker, K.; Lie, Y.T.; Pospieszczyk, A.; Rusbueldt, D.; Schweer, B.; Yu, Y.J.
1989-01-01
Investigations of the plasma edge in TEXTOR are presented on the one hand by comparing results obtained with poloidal and toroidal limiters and on the other hand by discussing general problems of plasma edge physics which are independent of the limiter configuration. The characteristic properties of plasma flow to the different limiters are analyzed and show e.g. that the fraction of total ion flow to the limiter is much larger in the case of a toroidal limiter (80%). Density and heat flux profiles are presented which demonstrate that for both types of limiters a significant steepening of the scrape-off layer (SOL) occurs close to the limiter, leading to a small heat load e-folding length of 5-8 mm. The velocity distribution of recycled neutral hydrogen at a main limiter has been determined from the Doppler broadening of the H α line. The data clearly show that a large fraction of particles (30-50%) is reflected at the limiter surface having energies of about the sheath potential. Significant isotopic effects (H/D) concerning the plasma edge properties and the plasma core are presented and their relation to enhanced particle and energy transport in hydrogen compared to deuterium is discussed. A decrease of the cross field diffusion coefficient with increasing density can be deduced from density profile measurements in the SOL and a comparison with density fluctuations is given. The role of oxygen for impurity release is demonstrated. A new type of wall conditioning - boronization - is described, with two major improvements for quasi stationary conditions: reduction of oxygen and better density control. Best results with ICRH have been obtained under these conditions. (orig.)
Physics parameter space of tokamak ignition devices
International Nuclear Information System (INIS)
Selcow, E.C.; Peng, Y.K.M.; Uckan, N.A.; Houlberg, W.A.
1985-01-01
This paper describes the results of a study to explore the physics parameter space of tokamak ignition experiments. A new physics systems code has been developed to perform the study. This code performs a global plasma analysis using steady-state, two-fluid, energy-transport models. In this paper, we discuss the models used in the code and their application to the analysis of compact ignition experiments. 8 refs., 8 figs., 1 tab
Kinetic modelling of runaway electron avalanches in tokamak plasmas.
Czech Academy of Sciences Publication Activity Database
Nilsson, E.; Decker, J.; Peysson, Y.; Granetz, R.S.; Saint-Laurent, F.; Vlainic, Milos
2015-01-01
Roč. 57, č. 9 (2015), č. článku 095006. ISSN 0741-3335 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : plasma physics * runaway electrons * knock-on collisions * tokamak * Fokker-Planck * runaway avalanches Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.404, year: 2015
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
Start of the international tokamak physics activity
International Nuclear Information System (INIS)
Campbell, D.
2001-01-01
This newsletter comprises a summary on the start of the International Tokamak Physics activity (ITPA) by Dr. D. Campbell, Chair of the ITPA Co-ordinating Committee. As the ITER EDA drew to a close, it became clear that it was desirable to establish a new mechanism in order to promote the continued development of the physics basis for burning plasma experiments and to preserve the invaluable collaborations between the major international fusion communities which had been established through the ITER physics expert groups. As a result of the discussions of the representatives of the European Union, Japan, the Russian Federation and the United States the agreed principles for conducting the International Tokamak Physics Activity (ITPA) were elaborated and ITPA topical physics groups were organized
The physics of magnetic confinement configurations : Tokamak theory and experiment
International Nuclear Information System (INIS)
Robinson, D.C.
1982-01-01
Several aspects, both theoretical and experimental, in plasma physics are discussed. The problem of magnetic confinement in Tokamak devices is treated. A discussion on the history of the development and on the future problems to be solved in Tokamaks is made. (L.C.) [pt
Tokamak Physics Experiment (TPX) power supply design and development
International Nuclear Information System (INIS)
Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.
1995-01-01
The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes
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
Generation of plasma rotation by ICRH in tokamaks
International Nuclear Information System (INIS)
Chang, C.; Phillips, C.K.; White, R.B.; Zweben, S.; Bonoli, P.T.; Rice, J.; Greenwald, M.; Grassie, J.S. de
2001-01-01
A physical mechanism to generate plasma rotation by ICRH is presented in a tokamak geometry. By breaking the omnigenity of resonant ion orbits, ICRH can induce a non-ambipolar minor-radial flow of resonant ions. This induces a return current j p r in the plasma, which then drives plasma rotation through the j p r xB force. It is estimated that the fast-wave power in the present-day tokamak experiments can be strong enough to give a significant modification to plasma rotation. (author)
Simplified models for radiational losses calculating a tokamak plasma
International Nuclear Information System (INIS)
Arutiunov, A.B.; Krasheninnikov, S.I.; Prokhorov, D.Yu.
1990-01-01
To determine the magnitudes and profiles of radiational losses in a Tokamak plasma, particularly for high plasma densities, when formation of MARFE or detached-plasma takes place, it is necessary to know impurity distribution over the ionization states. Equations describing time evolution of this distribution are rather cumbersome, besides that, transport coefficients as well as rate constants of the processes involving complex ions are known nowadays with high degree of uncertainty, thus it is believed necessary to develop simplified, half-analytical models describing time evolution of the impurities analysis of physical processes taking place in a Tokamak plasma on the base of the experimental data. (author) 6 refs., 2 figs
Numerical simulation of edge plasma in tokamak
International Nuclear Information System (INIS)
Chen Yiping; Qiu Lijian
1996-02-01
The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)
International Nuclear Information System (INIS)
Nuehrenberg, J.
1986-01-01
These proceedings contain the articles presented at the named conference. These concern numerical methods for astrophysical plasmas, the numerical simulation of reversed-field pinch dynamics, methods for numerical simulation of ideal MHD stability of axisymmetric plasmas, calculations of the resistive internal m=1 mode in tokamaks, parallel computing and multitasking, particle simulation methods in plasma physics, 2-D Lagrangian studies of symmetry and stability of laser fusion targets, computing of rf heating and current drive in tokamaks, three-dimensional free boundary calculations using a spectral Green's function method, as well as the calculation of three-dimensional MHD equilibria with islands and stochastic regions. See hints under the relevant topics. (HSI)
Simulation models for tokamak plasmas
International Nuclear Information System (INIS)
Dimits, A.M.; Cohen, B.I.
1992-01-01
Two developments in the nonlinear simulation of tokamak plasmas are described: (A) Simulation algorithms that use quasiballooning coordinates have been implemented in a 3D fluid code and a 3D partially linearized (Δf) particle code. In quasiballooning coordinates, one of the coordinate directions is closely aligned with that of the magnetic field, allowing both optimal use of the grid resolution for structures highly elongated along the magnetic field as well as implementation of the correct periodicity conditions with no discontinuities in the toroidal direction. (B) Progress on the implementation of a likeparticle collision operator suitable for use in partially linearized particle codes is reported. The binary collision approach is shown to be unusable for this purpose. The algorithm under development is a complete version of the test-particle plus source-field approach that was suggested and partially implemented by Xu and Rosenbluth
Advanced probes for edge plasma diagnostics on the CASTOR tokamak
Czech Academy of Sciences Publication Activity Database
Stöckel, Jan; Adámek, Jiří; Balan, P.; Hronová-Bilyková, Olena; Brotánková, Jana; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Horáček, Jan; Ionita, C.; Kocan, M.; Martines, E.; Pánek, Radomír; Peleman, P.; Schrittwieser, R.; Van Oost, G.; Žáček, František
2006-01-01
Roč. 63, č. 0 (2006), 012001-012002 E-ISSN 1742-6596. [SECOND INTERNATIONAL WORKSHOP AND SUMMER SCHOOL ON PLASMA PHYSICS. Kiten, 03.07.2006-09.07.2006] R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma * tokamak * electric probes * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics
International Nuclear Information System (INIS)
Budny, R.V.; Alper, B.; Borba, D.; Cordey, J.G.; Ernst, D.R.; Gowers, C.
2001-01-01
First results of gyrokinetic analysis of JET [Joint European Torus] ELMy [Edge Localized Modes] H-mode [high-confinement modes] plasmas are presented. ELMy H-mode plasmas form the basis of conservative performance predictions for tokamak reactors of the size of ITER [International Thermonuclear Experimental Reactor]. Relatively high performance for long duration has been achieved and the scaling appears to be favorable. It will be necessary to sustain low Z(subscript eff) and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one with an intense gas puff and the other with a spontaneous transition between Type I to III ELMs at the heating power threshold. Linear gyrokinetic analysis gives the growth rate, gamma(subscript lin) of the fastest growing modes. The flow-shearing rate omega(subscript ExB) and gamma(subscript lin) are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high |gamma(subscript ExB)|/gamma(subscript lin) near the top of the pedestal for high confinement
Plasma equilibrium and instabilities in tokamaks
International Nuclear Information System (INIS)
Caldas, I.L.; Vannucci, A.
1985-01-01
A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt
Discontinuous Galerkin methods for plasma physics in the scrape-off layer of tokamaks
International Nuclear Information System (INIS)
Michoski, C.; Meyerson, D.; Isaac, T.; Waelbroeck, F.
2014-01-01
A new parallel discontinuous Galerkin solver, called ArcOn, is developed to describe the intermittent turbulent transport of filamentary blobs in the scrape-off layer (SOL) of fusion plasma. The model is comprised of an elliptic subsystem coupled to two convection-dominated reaction–diffusion–convection equations. Upwinding is used for a class of numerical fluxes developed to accommodate cross product driven convection, and the elliptic solver uses SIPG, NIPG, IIPG, Brezzi, and Bassi–Rebay fluxes to formulate the stiffness matrix. A novel entropy sensor is developed for this system, designed for a space–time varying artificial diffusion/viscosity regularization algorithm. Some numerical experiments are performed to show convergence order on manufactured solutions, regularization of blob/streamer dynamics in the SOL given unstable parameterizations, long-time stability of modon (or dipole drift vortex) solutions arising in simulations of drift-wave turbulence, and finally the formation of edge mode turbulence in the scrape-off layer under turbulent saturation conditions
Tokamak-FED plasma-engineering assessments
International Nuclear Information System (INIS)
Peng, Y.K.M.; Lyon, J.F.; Rutherford, P.H.
1981-01-01
A wide range of plasma assumptions and scenarios has been examined for the current US tokamak FED concept, which aims to provide a controlled, long pulse (approx. 100 s) burning plasma with an energy amplification of greater than or equal to 5, a fusion power of 180 MW, and a neutron wall load of greater than or equal to 0.4 MW/m 2 . The results of the assessment suggest that the current FED baseline parameters of R = 5.0 m, B/sub t/ = 3.6 T, a = 1.3 m, b = 2.1 m (D-shape), and I/sub p/ = 5.4 MA are appropriate in reaching the above plasma performance, despite uncertainties in several plasma physics areas, such as confinement scaling, achievable beta, impurity control, etc. To enhance the probability of achieving fusion ignition and to provide some margin against a short fall in our physics projections in FED, a limited operating capability at B/sub t/ = 4.6 T and I/sub p/ = 6.5 MA is incorporated. Various other options and remedies have also been assessed aiming to alleviate the impact of the uncertainties on the FED design concept. These approaches appear promising because they can be studied within the current fusion physics program and may lead to drastically more cost-effective FED concepts
Farahani, N Darestani; Davani, F Abbasi
2015-10-01
This investigation is about plasma modeling for the control of vertical instabilities in Damavand tokamak. This model is based on online magnetic measurement. The algebraic equation defining the vertical position in this model is based on instantaneous force-balance. Two parameters in this equation, including decay index, n, and lambda, Λ, have been considered as functions of time-varying poloidal field coil currents and plasma current. Then these functions have been used in a code generated for modeling the open loop response of plasma. The main restriction of the suitability analysis of the model is that the experiments always have to be performed in the presence of a control loop for stabilizing vertical position. As a result, open loop response of the system has been identified from closed loop experimental data by nonlinear neural network identification method. The results of comparison of physical model with identified open loop response from closed loop experiments show root mean square error percentage less than 10%. The results are satisfying that the physical model is useful as a Damavand tokamak vertical movement simulator.
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 physics of an ignited tokamak
International Nuclear Information System (INIS)
Troyon, F.
1990-10-01
There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab
A programmatic framework for the Tokamak Physics Experiment (TPX)
International Nuclear Information System (INIS)
Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.
1993-01-01
Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program
Plasma diagnostics for the compact ignition tokamak
International Nuclear Information System (INIS)
Medley, S.S.; Young, K.M.
1988-06-01
The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10 21 m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs
Plasma Sprayed Tungsten-based Coatings and their Usage in Edge Plasma Region of Tokamaks
Czech Academy of Sciences Publication Activity Database
Matějíček, Jiří; Weinzettl, Vladimír; Dufková, Edita; Piffl, Vojtěch; Peřina, Vratislav
2006-01-01
Roč. 51, č. 2 (2006), s. 179-191 ISSN 0001-7043 Grant - others:Evropská unie EFDA Task TW-5-TVM-PSW (EU – Euratom) Institutional research plan: CEZ:AV0Z20430508; CEZ:AV0Z10480505 Keywords : plasma sprayed coatings * fusion * plasma facing components * tungsten * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics
Abstracts of 7th Ukrainian conference on controlled nuclear fusion and plasma physics
International Nuclear Information System (INIS)
1999-01-01
This conference discussed the main directions of plasma physics development in Ukraine. The experimental and theoretical research on stellarators and theoretical results of physical processes in tokamak plasma studied. The investigation of spherical tokamaks were plasma physics began
Plasma diagnostics using synchrotron radiation in tokamaks
International Nuclear Information System (INIS)
Fidone, I.; Giruzzi, G.; Granata, G.
1995-09-01
This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs
Increase in beta limit in tokamak plasmas
International Nuclear Information System (INIS)
Kamada, Yutaka
2003-01-01
This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)
Plasma-gun fueling for tokamak reactors
International Nuclear Information System (INIS)
Ehst, D.A.
1980-11-01
In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment
Edge plasma diagnostics on Tore Supra tokamak
International Nuclear Information System (INIS)
Fujita, Junji
1991-01-01
From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)
Physics design of an ultra-long pulsed tokamak reactor
International Nuclear Information System (INIS)
Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.
1993-01-01
A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab
Energy Technology Data Exchange (ETDEWEB)
Sarazin, Y
2004-03-01
This document gathers the lectures made in the framework of a Ph.D level physics class dedicated to plasma physics. This course is made up of 3 parts : 1) collisions and transport, 2) transport and turbulence, and 3) study of a few exchange instabilities. More precisely the first part deals with the following issues: thermonuclear fusion, Coulomb collisions, particles trajectories in a tokamak, neo-classical transport in tokamaks, the bootstrap current, and ware pinch. The second part involves: particle transport in tokamaks, quasi-linear transport, resonance islands, resonance in tokamaks, from quasi to non-linear transport, and non-linear saturation of turbulence. The third part deals with: shift velocities in fluid theory, a model for inter-change instabilities, Rayleigh-Benard instability, Hasegawa-Wakatani model, and Hasegawa-Mima model. This document ends with a series of appendices dealing with: particle-wave interaction, determination of the curvature parameter G, Rossby waves.
Tokamak nonmaxwellian plasma dynamics in thermonuclear regime
International Nuclear Information System (INIS)
Cotsaftis, M.
1987-01-01
To reach ignition in a Tokamak plasma, large additional power P aux has to be injected in the device on top of the Joule heating P OH =VI r , V the plasma loop voltage, I r the resistive port of plasma current. Typi-cally JH ∼ 1 KeV, whereas ignition would requi- re IG ∼ 7-10 KeV. To gain this factor 7, one at least should inject additional power P aux ∼ 7P OH , supposing that nothing, especially the heat transport, is modified. This is by far not the case, with the so-called energy lifetime degradation, largely observed in oil experiments (but less dramatic with divertors), where energy lifetime tau E behaves like P tot -b with b∼1/2. In large machines where ignition temperature is the target to be imperiously reached, this implies to inject a very large power, typically P aux ∼ 50 to 100 MW, depending on size and parameters and on actual transport. So it is of importance with such figures, or even larger ones owing to uncertain ties, to optimize at best injected power by increasing its efficiency, both with respect to possible transport laws, and to physical phenomena governing heat flow in the system from the sources. This leads to the concept of scenarios, as time sequences of power input, where physical properties of the plasma system are used to build up ion temperature so that ignition is reached with minimum P tot = P OH + P aux and with fixed Q = Q o > 1. Elements for this study are given. The method is outlined. The resulting system of equations describing the evolution of a thermonuclear plasma is given
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
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)
A complex probe for tokamak plasma edge conditions
International Nuclear Information System (INIS)
Castro, R.M. de; Silva, R.P. da; Heller, M.V.A.P.; Caldas, I.L.; Nascimento, I.C.; Degasperi, F.T.
1995-01-01
The study of the physical processes that occur in the plasma edge of tokamak machines has recently grown due to the evidence that these processes influence those that occur in the center of the plasma column. Experimental studies show the existence of a strong level of fluctuations in the plasma edge. The results of these studies indicate that these fluctuations enhance particle and energy transport and degrade the confinement. In order to investigate these processes in the plasma edge of the TBR-1 Tokamak, a Langmuir probe array, a triple and a set of magnetic probes have been designed and constructed. With this set probes the mean and fluctuation values of the magnetic field were detected and correlated with the fluctuating parameters obtained with the electrostatic probes. (author). 7 refs., 5 figs
Plasma position control in SST1 tokamak
Indian Academy of Sciences (India)
also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...
INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
International Nuclear Information System (INIS)
HUMPHREYS, D.A.; FERRON, J.R.; JOHNSON, R.D; LEUER, J.A.; PENAFLOR, B.G.; WALKER, M.L.; WELANDER, A.S.; KHAYRUTDINOV, R.R; DOKOUKA, V.; EDGELL, D.H.; FRANSSON, C.M.
2004-03-01
OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance
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
Electron cyclotron heating (ECH) of tokamak plasmas
International Nuclear Information System (INIS)
Hoshino, Katsumichi
1990-01-01
Electron cyclotron heating (ECH) is one of the intense methods of plasma heating, and which utilizes the collisionless electron-cyclotron-resonance-interaction between the launched electromagnetic waves (called electron cyclotron waves) and electrons which are one of the constituents of the high temperature plasmas. Another constituent, namely the ions which are subject to nuclear fusion, are heated indirectly but strongly and instantly (in about 0.1 s) by the collisions with the ECH-heated electrons in the fusion plasmas. The recent progress on the development of high-power and high-frequency millimeter-wave-source enabled the ECH experiments in the middle size tokamaks such as JFT-2M (Japan), Doublet III (USA), T-10 (USSR) etc., and ECH has been demonstrated to be the sure and intense plasma heating method. The ECH attracts much attention for its remarkable capabilities; to produce plasmas (pre-ionization), to heat plasmas, to drive plasma current for the plasma confinement, and recently especially by the localization and the spatial controllability of its heating zone, which is beneficial for the fine controls of the profiles of plasma parameters (temperature, current density etc.), for the control of the magnetohydrodynamic instabilities, or for the optimization/improvement of the plasma confinement characteristics. Here, the present status of the ECH studies on tokamak plasmas are reviewed. (author)
Coherent structures in tokamak plasmas workshop: Proceedings
International Nuclear Information System (INIS)
Koniges, A.E.; Craddock, G.G.
1992-08-01
Coherent structures have the potential to impact a variety of theoretical and experimental aspects of tokamak plasma confinement. This includes the basic processes controlling plasma transport, propagation and efficiency of external mechanisms such as wave heating and the accuracy of plasma diagnostics. While the role of coherent structures in fluid dynamics is better understood, this is a new topic for consideration by plasma physicists. This informal workshop arose out of the need to identify the magnitude of structures in tokamaks and in doing so, to bring together for the first time the surprisingly large number of plasma researchers currently involved in work relating to coherent structures. The primary purpose of the workshop, in addition to the dissemination of information, was to develop formal and informal collaborations, set the stage for future formation of a coherent structures working group or focus area under the heading of the Tokamak Transport Task Force, and to evaluate the need for future workshops on coherent structures. The workshop was concentrated in four basic areas with a keynote talk in each area as well as 10 additional presentations. The issues of discussion in each of these areas was as follows: Theory - Develop a definition of structures and coherent as it applies to plasmas. Experiment - Review current experiments looking for structures in tokamaks, discuss experimental procedures for finding structures, discuss new experiments and techniques. Fluids - Determine how best to utilize the resource of information available from the fluids community both on the theoretical and experimental issues pertaining to coherent structures in plasmas. Computation - Discuss computational aspects of studying coherent structures in plasmas as they relate to both experimental detection and theoretical modeling
Controlled fusion and plasma physics
International Nuclear Information System (INIS)
1995-01-01
This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC)
Controlled fusion and plasma physics
Energy Technology Data Exchange (ETDEWEB)
NONE
1996-12-31
This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).
Controlled fusion and plasma physics
Energy Technology Data Exchange (ETDEWEB)
NONE
1995-12-31
This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).
Analysis of tokamak plasma confinement modes using the fast
Indian Academy of Sciences (India)
The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...
The division of plasma physics
International Nuclear Information System (INIS)
Evans, T.E.; Guilhem, D.; Klepper, C.C.
1990-07-01
The investigations presented in the 31th meeting on plasma physics were: the main results and observations during the ergodic divertor experiments in Tore Supra tokamak; the modifications of power scrape-off-length and power deposition during various configurations in Tore Supra plasmas; the results of pressure measurements and particle fluxes in the Tore Supra pump limiter
Economic trends of tokamak power plants independent of physics scaling models
International Nuclear Information System (INIS)
Reid, R.L.; Steiner, D.
1978-01-01
This study examines the effects of plasma radius, field on axis, plasma impurity level, and aspect ratio on power level and unit capital cost, $/kW/sub e/, of tokamak power plants sized independent of plasma physics scaling models. It is noted that tokamaks sized in this manner are thermally unstable based on trapped particle scaling relationships. It is observed that there is an economic advantage for larger power level tokamaks achieved by physics independent sizing; however, the incentive for increased power levels is less than that for fission reactors. It is further observed that the economic advantage of these larger power level tokamaks is decreased when plasma thermal stability measures are incorporated, such as by increasing the plasma impurity concentration. This trend of economy with size obtained by physics independent sizing is opposite to that observed when the tokamak designs are constrained to obey the trapped particle and empirical scaling relationships
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.
Drummond, James E
1961-01-01
A historic snapshot of the field of plasma physics, this fifty-year-old volume offers an edited collection of papers by pioneering experts in the field. In addition to assisting students in their understanding of the foundations of classical plasma physics, it provides a source of historic context for modern physicists. Highly successful upon its initial publication, this book was the standard text on plasma physics throughout the 1960s and 70s.Hailed by Science magazine as a ""well executed venture,"" the three-part treatment ranges from basic plasma theory to magnetohydrodynamics and microwa
Advanced tokamak physics in DIII-D
Energy Technology Data Exchange (ETDEWEB)
Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)
2000-12-01
Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)
Properties of the tokamak edge plasma
International Nuclear Information System (INIS)
Wolff, H.
1988-01-01
A short review of some features of the edge plasma in limiter tokamaks is given. The limits of the simple one-dimensional scrape-off layer (SOL) model and the relation between the core plasma are discussed. Multifaceted asymmetric radiation from the edge (MARFE) phenomena and detached plasma are closely connected with the particle and energy balance of the SOL. Their occurrence is based on the relation of plasma parameters of the edge plasma to those of the core. Important problems of plasma wall interactions are the detection of the impurity sources and sinks and the study of the impurity transport and shielding. The non-uniform character of plasma wall interactions and their dependence on the discharge performance still renders difficult any theoretical forecast of impurity distribution and transport and calls for better diagnostics. (author)
Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions
International Nuclear Information System (INIS)
Sengoku, Seio
1985-08-01
Relation between plasma-wall interactions and confinement characteristics of a tokamak plasma with respect to both impurity and fuel particle controls is discussed. Following results are obtained from impurity control studies: (1) Ion sputtering is the dominant mechanism of impurity release in a steady state tokamak discharge. (2) By applying carbon coating on entire first wall of DIVA tokamak, dominant radiative region is concentrated more in boundary plasma resulting a hot peripheral plasma with cold boundary plasma. (3) A physical model of divertor functions about impurity control is empilically obtained. By a computer simulation based on above model with respect to divertor functions for JT-60 tokamak, it is found that the allowable electron temperature of the divertor plasma is not restricted by a condition that the impurity release due to ion sputtering does not increase continuously. (4) Dense and cold divertor plasma accompanied with strong remote radiative cooling was diagnosed along the magnetic field line in the simple poloidal divertor of DOUBLET III tokamak. Strong particle recycling region is found to be localized near the divertor plate. by and from particle control studies: (1) The INTOR scaling on energy confinement time is applicable to high density region when a core plasma is fueled directly by solid deuterium pellet injection in DOUBLET III tokamak. (2) As remarkably demonstrated by direct fueling with pellet injection, energy confinement characteristics can be improved at high density range by decreasing particle deposition at peripheral plasma in order to reduce plasma-wall interaction. (3) If the particle deposition at boundary layer is necessarily reduced, the electron temperature at the boundary or divertor region increases due to decrease of the particle recycling and the electron density there. (J.P.N.)
Physics evaluation of compact tokamak ignition experiments
International Nuclear Information System (INIS)
Uckan, N.A.; Houlberg, W.A.; Sheffield, J.
1985-01-01
At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/ 2 /q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs
Physics aspects of the Compact Ignition Tokamak
International Nuclear Information System (INIS)
Post, D.; Bateman, G.; Houlberg, W.
1986-11-01
The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha-particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve the condition of ntau/sub E/ ∼ 2 x 10 20 sec m -3 required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which improves the energy confinement, and provides a high level of ohmic heating. The present CIT design also has a gigh degree of elongation (k ∼ 1.8) to aid in producing the large plasma current. A double null poloidal divertor and a pellet injector are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Since auxiliary heating is expected to be necessary to achieve ignition, 10 to 20 MW of Ion Cyclotron Radio Frequency (ICRF) is to be provided
Plasma features and alpha particle transport in low-aspect ratio tokamak reactor
International Nuclear Information System (INIS)
Xu Qiang; Wang Shaojie
1997-06-01
The results of the experiment and theory from low-aspect ratio tokamak devices have proved that the MHD stability will be improved. Based on present plasma physics and extrapolation to reduced aspect ratio, the feature of physics of low-aspect ratio tokamak reactor is discussed primarily. Alpha particle confinement and loss in the self-justified low-aspect ratio tokamak reactor parameters and the effect of alpha particle confinement and loss for different aspect ratio are calculated. The results provide a reference for the feasible research of compact tokamak reactor. (9 refs., 2 figs., 3 tabs.)
Dynamics of the edge transport barrier at plasma biasing on the CASTOR tokamak
Czech Academy of Sciences Publication Activity Database
Stöckel, Jan; Spolaore, M.; Peleman, P.; Brotánková, Jana; Horáček, Jan; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Kocan, M.; Martines, E.; Pánek, Radomír; Sharma, A.; Van Oost, G.
2006-01-01
Roč. 12, č. 6 (2006), s. 19-23 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * transport barrier * relaxations Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html
Czech Academy of Sciences Publication Activity Database
Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Krämer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Peleman, P.; Razumova, K.; Stöckel, Jan; Vershkov, V.; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Dnestrovskij, A.Yu.; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Jachmin, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Soldatov, S.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Zimmermann, O.
2006-01-01
Roč. 12, č. 6 (2006), s. 14-19 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * improved confinement * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html
Lower hybrid current drive in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1999-03-01
Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar{sub e} - 10{sup 20}m{sup -3}, ALCATOR-C) and the highest current drive efficiency ({eta}{sub CD} = 3.5x10{sup 19} m{sup -2}A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)
Lower hybrid current drive in tokamak plasmas
International Nuclear Information System (INIS)
Ushigusa, Kenkichi
1999-03-01
Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar e - 10 20 m -3 , ALCATOR-C) and the highest current drive efficiency (η CD = 3.5x10 19 m -2 A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)
Scrape-off layer tokamak plasma turbulence
Bisai, N.; Singh, R.; Kaw, P. K.
2012-05-01
Two-dimensional (2D) interchange turbulence in the scrape-off layer of tokamak plasmas and their subsequent contribution to anomalous plasma transport has been studied in recent years using electron continuity, current balance, and electron energy equations. In this paper, numerically it is demonstrated that the inclusion of ion energy equation in the simulation changes the nature of plasma turbulence. Finite ion temperature reduces floating potential by about 15% compared with the cold ion temperature approximation and also reduces the radial electric field. Rotation of plasma blobs at an angular velocity about 1.5×105 rad/s has been observed. It is found that blob rotation keeps plasma blob charge separation at an angular position with respect to the vertical direction that gives a generation of radial electric field. Plasma blobs with high electron temperature gradients can align the charge separation almost in the radial direction. Influence of high ion temperature and its gradient has been presented.
Magnetohydrodynamic stability of tokamak edge plasmas
International Nuclear Information System (INIS)
Connor, J.W.; Hastie, R.J.; Wilson, H.R.; Miller, R.L.
1998-01-01
A new formalism for analyzing the magnetohydrodynamic stability of a limiter tokamak edge plasma is developed. Two radially localized, high toroidal mode number n instabilities are studied in detail: a peeling mode and an edge ballooning mode. The peeling mode, driven by edge current density and stabilized by edge pressure gradient, has features which are consistent with several properties of tokamak behavior in the high confinement open-quotes Hclose quotes-mode of operation, and edge localized modes (or ELMs) in particular. The edge ballooning mode, driven by the pressure gradient, is identified; this penetrates ∼n 1/3 rational surfaces into the plasma (rather than ∼n 1/2 , expected from conventional ballooning mode theory). Furthermore, there exists a coupling between these two modes and this coupling provides a picture of the ELM cycle
High beta plasmas in the PBX tokamak
International Nuclear Information System (INIS)
Bol, K.; Buchenauer, D.; Chance, M.
1986-04-01
Bean-shaped configurations favorable for high β discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present β limit
Physics issues of high bootstrap current tokamaks
International Nuclear Information System (INIS)
Ozeki, T.; Azumi, M.; Ishii, Y.
1997-01-01
Physics issues of a tokamak plasma with a hollow current profile produced by a large bootstrap current are discussed based on experiments in JT-60U. An internal transport barrier for both ions and electrons was obtained just inside the radius of zero magnetic shear in JT-60U. Analysis of the toroidal ITG microinstability by toroidal particle simulation shows that weak and negative shear reduces the toroidal coupling and suppresses the ITG mode. A hard beta limit was observed in JT-60U negative shear experiments. Ideal MHD mode analysis shows that the n = 1 pressure-driven kink mode is a plausible candidate. One of the methods to improve the beta limit against the kink mode is to widen the negative shear region, which can induce a broader pressure profile resulting in a higher beta limit. The TAE mode for the hollow current profile is less unstable than that for the monotonic current profile. The reason is that the continuum gaps near the zero shear region are not aligned when the radius of q min is close to the region of high ∇n e . Finally, a method for stable start-up for a plasma with a hollow current profile is describe, and stable sustainment of a steady-state plasma with high bootstrap current is discussed. (Author)
Anisotropic plasma with flows in tokamak: Steady state and stability
International Nuclear Information System (INIS)
Ilgisonis, V.I.
1996-01-01
An adequate description of equilibrium and stability of anisotropic plasma with macroscopic flows in tokamaks is presented. The Chew-Goldberger-Low (CGL) approximation is consistently used to analyze anisotropic plasma dynamics. The admissible structure of a stationary flow is found to be the same as in the ideal magnetohydrodynamics with isotropic pressure (MHD), which means an allowance for the same relabeling symmetry as in ideal MHD systems with toroidally nested magnetic surfaces. A generalization of the Grad-Shafranov equation for the case of anisotropic plasma with flows confined in the axisymmetric magnetic field is derived. A variational principle was obtained, which allows for a stability analysis of anisotropic pressure plasma with flows, and takes into account the conservation laws resulting from the relabeling symmetry. This principle covers the previous stability criteria for static CGL plasma and for ideal MHD flows in isotropic plasma as well. copyright 1996 American Institute of Physics
Plasma physics for controlled fusion
International Nuclear Information System (INIS)
Miyamoto, K.
2010-01-01
The primary objective of this lecture note is to present the theories and experiments of plasma physics for recent activities of controlled fusion research for graduate and senior undergraduate students. Chapters 1-6 describe the basic knowledge of plasma and magnetohydrodynamics (MHD). MHD instabilities limit the beta ratio (ratio of plasma pressure to magnetic pressure) of confined plasma. Chapters 7-9 provide the kinetic theory of hot plasma and discuss the wave heating and non-inductive current drive. The dispersion relation derived by the kinetic theory are used to discuss plasma waves and perturbed modes. Landau damping is the essential mechanism of plasma heating and the stabilization of perturbation. Landau inverse damping brings the amplification of waves and the destabilization of perturbed modes. Chapter 10 explains the plasma transport due to turbulence, which is the most important and challenging subject for plasma confinement. Theories and simulations including subject of zonal flow are introduced. Chapters 11, 12 and 13 describe the recent activities of tokamak including ITER as well as spherical tokamak, reversed field pinch (RFP) and stellarator including quasi-symmetric configurations. Emphasis has been given to tokamak research since it made the most remarkable progress and the construction phase of 'International Tokamak Experimental Reactor' called ITER has already started. (author)
Plasma engineering analyses of tokamak reactor operating space
International Nuclear Information System (INIS)
Houlberg, W.; Attenberger, S.E.
1981-01-01
A comprehensive method is presented for analyzing the potential physics operating regime of fusion reactor plasmas with detailed transport codes. Application is made to the tokamak Fusion Engineering Device (FED). The relationships between driven and ignited operation and supplementary heating requirements are examined. The reference physics models give a finite range of density and temperature over which physics objectives can be reached. Uncertainties in the confinement scaling and differences in supplementary heating methods can expand or contract this operating regime even to the point of allowing ignition with the more optimistic models
Relativistic runaway electrons in tokamak plasmas
International Nuclear Information System (INIS)
Jaspers, R.E.
1995-01-01
Runaway electrons are inherently present in a tokamak, in which an electric field is applied to drive a toroidal current. The experimental work is performed in the tokamak TEXTOR. Here runaway electrons can acquire energies of up to 30 MeV. The runaway electrons are studied by measuring their synchrotron radiation, which is emitted in the infrared wavelength range. The studies presented are unique in the sense that they are the first ones in tokamak research to employ this radiation. Hitherto, studies of runaway electrons revealed information about their loss in the edge of the discharge. The behaviour of confined runaways was still a terra incognita. The measurement of the synchrotron radiation allows a direct observation of the behaviour of runaway electrons in the hot core of the plasma. Information on the energy, the number and the momentum distribution of the runaway electrons is obtained. The production rate of the runaway electrons, their transport and the runaway interaction with plasma waves are studied. (orig./HP)
Evaluation of the plasma parameters in COMPASS tokamak divertor area
Czech Academy of Sciences Publication Activity Database
Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud
2012-01-01
Roč. 356, č. 1 (2012), s. 012007 ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf
Turbulence studies in tokamak boundary plasmas with realistic divertor geometry
International Nuclear Information System (INIS)
Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.
2001-01-01
Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)
Turbulence studies in tokamak boundary plasmas with realistic divertor geometry
International Nuclear Information System (INIS)
Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.
1999-01-01
Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)
Turbulence and abnormal transport in tokamak plasmas
International Nuclear Information System (INIS)
Garbet, X.
1988-09-01
Microinstabilities in linear and nonlinear tokamak plasmas were studied. A variational method based on the existence of a system of angular variables and action for the charged particles in the magnetic configuration of a tokamak is described. The corresponding functional, extremal in relation to the fluctuating electromagnetic field, is calculated analytically, taking into account the effects of the toroidal geometry. A numerical code, TORRID, was derived from these principles and the main instabilities, especially ion instabilities and microtearing, were studied linearly. Nonlinear methods were also applied to microtearing. Quasi-linear transport coefficients are derived from a principle of minimum entropy production. Thermal ionic conductivity and viscosity are calculated for an ionic turbulence [fr
Turbulence and abnormal transport in tokamak plasmas
International Nuclear Information System (INIS)
Garbet, X.
1988-06-01
The objective of this thesis is the study of plasma microinstabilities in linear and nonlinear tokamak regime. After a brief review of experimental results the theoretical tools used in this study are presented. A variational method founded on the existence of angular variables system and on action for charged particles in tokamak configurations is detailed. The correspondent functional extreme with regard to fluctuating electromagnetic field, is calculated analytically with taking into account the toroidal geometry. A numerical code, TORRID, has been constructed on this principle and the main instabilities, particularly ionic instabilities and microtearing, has been linearly studied. The most simple non linear methods are rewieved and applied at the microtearing instabilities. The quasilinear transport coefficients are deducted of an entropy minimum production principle. The ionic thermic conductivity and the viscosity are calculated for an ionic turbulence [fr
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
Plasma transport in a compact ignition tokamak
International Nuclear Information System (INIS)
Singer, C.E.; Ku, L.P; Bateman, G.
1987-02-01
Nominal predicted plasma conditions in a compact ignition tokamak are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models which have given almost equally good fits to experimental data. Using a transport model which best fits the data, thermonuclear ignition occurs in a Compact Ignition Tokamak design with major radius 1.32 m, plasma half-width 0.43 m, elongation 2.0, and toroidal field and plasma current ramped in six seconds from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the 3 He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates, have a large effect on ignition and on the maximum beta that can be achieved
Neutron measurement techniques for tokamak plasmas
International Nuclear Information System (INIS)
Jarvis, O.N.
1994-01-01
The present article reviews the neutron measurement techniques that are currently being applied to the study of tokamak plasmas. The range of neutron energies of primary interest is limited to narrow bands around 2.5 and 14 MeV, and the variety of measurements that can be made for plasma diagnostic purposes is also restricted. To characterize the plasma as a neutron source, it is necessary only to measure the total neutron emission, the relative neutron emissivity as a function of position throughout the plasma, and the energy spectra of the emitted neutrons. In principle, such measurements might be expected to be relatively easy. That this is not the case is, in part, attributable to practical problems of accessibility to a harsh environment but is mostly a consequence of the time-scale on which the measurements have to be made and of the wide range of neutron emission intensities that have to be covered: for tokamak studies, the time-scale is of the order of 1 to 100 ms and the neutron intensity ranges from 10 12 to 10 19 s -1 . (author)
Mathematical modeling plasma transport in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)
1997-01-01
In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10^{20}/m^{3} with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.
Mathematical modeling plasma transport in tokamaks
International Nuclear Information System (INIS)
Quiang, Ji
1995-01-01
In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%
Effects of orbit squeezing on neoclassical toroidal plasma viscosity in tokamaks
Czech Academy of Sciences Publication Activity Database
Shaing, K.C.; Sabbagh, S.A.; Chu, M.S.; Bécoulet, M.; Cahyna, Pavel
2008-01-01
Roč. 15, č. 8 (2008), 082505-1-082505-8 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma instability * plasma magnetohydrodynamics * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2965146
Collisional boundary layer analysis for neoclassical toroidal plasma viscosity in tokamaks
Czech Academy of Sciences Publication Activity Database
Shaing, K.C.; Cahyna, Pavel; Bécoulet, M.; Park, J.-K.; Sabbagh, S.A.; Chu, M.S.
2008-01-01
Roč. 15, č. 8 (2008), 082506-1-7 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2969434
Multi-field plasma sandpile model in tokamaks and applications
Peng, X. D.; Xu, J. Q.
2016-08-01
A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.
Energy Technology Data Exchange (ETDEWEB)
Colunga S, S
1990-07-15
In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)
Plasma startup patterns in tokamak reactors
International Nuclear Information System (INIS)
Maki, Koichi; Tone, Tatsuzo.
1983-01-01
Plasma startup patterns are studied from the viewpoint of net power loss represented by the total power loss less the α-particle heating power. The existence is shown of a critical temperature of plasma at which the net power loss becomes independent of plasma density. Observations are made which indicate that the net power loss decreases with lowering plasma density in the range below the critical temperature and vice versa, whether governed by empirical or trapped-ion scaling laws. A startup pattern is presented which minimizes the net power loss during startup, and which prescribes that: (1) The plasma density should be kept as low as possible until the plasma is heated up to the critical temperature; (2) thereafter, the plasma density should be increased to its steady state value while retaining the critical temperature; and (3) finally, with the density kept constant, the temperature should be further raised to its steady state value. The net power loss at critical temperature represents the lower limit of heating power required to bring the plasma to steady state in tokamak reactors. (author)
Thermonuclear Tokamak plasmas in the presence of fusion alpha particles
International Nuclear Information System (INIS)
Anderson, D.; Hamnen, H.; Lisak, M.
1988-01-01
In this overview, we have focused on several results of the thermonuclear plasma research pertaining to the alpha particle physics and diagnostics in a fusion tokamak plasma. As regards the discussion of alpha particle effects, two distinct classes of phenomena have been distinguished: the simpler class containing phenomena exhibited by individual alpha particles under the influence of bulk plasma properties and, the more complex class including collective effects which become important for increasing alpha particle density. We have also discussed several possibilities to investigate alpha particle effects by simulation experiments using an equivalent population of highly energetic ions in the plasma. Generally, we find that the present theoretical knowledge on the role of fusion alpha particles in a fusion tokamak plasma is incomplete. There are still uncertainties and partial lack of quantitative results in this area. Consequently, further theoretical work and, as far a possible, simulation experiments are needed to improve the situation. Concerning the alpha particle diagnostics, the various diagnostic techniques and the status of their development have been discussed in two different contexts: the escaping alpha particles and the confined alpha particles in the fusion plasma. A general conclusion is that many of the different diagnostic methods for alpha particle measurements require further major development. (authors)
Physics design requirements for the Tokamak Physics Experiment (TPX)
International Nuclear Information System (INIS)
Neilson, G.H.; Goldston, R.J.; Jardin, S.C.; Reiersen, W.T.; Porkolab, M.; Ulrickson, M.
1993-01-01
The design of TPX is driven by physics requirements that follow from its mission. The tokamak and heating systems provide the performance and profile controls needed to study advanced steady state tokamak operating modes. The magnetic control systems provide substantial flexibility for the study of regimes with high beta and bootstrap current. The divertor is designed for high steady state power and particle exhaust
Turbulent ion heating in TCV Tokamak plasmas
International Nuclear Information System (INIS)
Schlatter, Ch.
2009-08-01
The Tokamak à configuration variable (TCV) features the highest electron cyclotron wave power density available to resonantly heat (ECRH) the electrons and to drive noninductive currents in a fusion grade plasma (ECCD). In more than 15 years of exploitation, much effort has been expended on real and velocity space engineering of the plasma electron energy distribution function and thus making electron physics a major research contribution of TCV. When a plasma was first subjected to ECCD, a surprising energisation of the ions, perpendicular to the confining magnetic field, was observed on the charge exchange spectrum measured with the vertical neutral particle analyser (VNPA). It was soon concluded that the ion acceleration was not due to power equipartition between electrons and ions, which, due to the absence of direct ion heating on TCV, has thus far been considered as the only mechanism heating the ions. However, although observed for more than ten years, little attention was paid to this phenomenon, whose cause has remained unexplained to date. The key subject of this thesis is the experimental study of this anomalous ion acceleration, the characterisation in terms of relevant parameters and the presentation of a model simulation of the potential process responsible for the appearance of fast ions. The installation of a new compact neutral particle analyser (CNPA) with an extended high energy range (≥ 50 keV) greatly improved the fast ion properties diagnosis. The CNPA was commissioned and the information derived from its measurement (ion temperature and density, isotopic plasma composition) was validated against other ion diagnostics, namely the active carbon charge exchange recombination spectroscopy system (CXRS) and a neutron counter. In ohmic plasmas, where the ion heating agrees with classical theory, the radial ion temperature profile was successfully reconstructed by vertically displacing the plasma across the horizontal CNPA line of sight. Active
Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas
J.W. Haverkort (Willem)
2013-01-01
htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma
Design of Tokamak plasma with high Tc superconducting coils
International Nuclear Information System (INIS)
Uchimoto, T.; Miya, K.; Yoshida, Y.; Yamada, T.
1999-01-01
This paper presents a design of tokamak plasma in light of how the small ignited tokamak is possible with use of the HTSC coils as plasma stabilizer. The same data base and formulas as ITER are here used and any innovative technology other than the HTSC stabilizing coils is not assumed. (author)
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
International Nuclear Information System (INIS)
Colunga S, S.
1990-07-01
In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)
Abstracts of the 23rd European physical society conference on controlled fusion and plasma physics
Energy Technology Data Exchange (ETDEWEB)
Goutych, I F; Gresillon, D; Sitenko, A G
1997-12-31
This document contains the abstracts of the invited and contributed papers presented at 23 EPS conference on controlled fusion and plasma physics. The main contents are: tokamaks, stellarators; alternative magnetic confinement; plasma edge physics; plasma heating and current drive; plasma diagnostics; basic collisionless plasma physics; high intensity laser produced plasmas and inertial confinement; low-temperature plasmas.
Abstracts of the 23rd European physical society conference on controlled fusion and plasma physics
International Nuclear Information System (INIS)
Goutych, I.F.; Gresillon, D.; Sitenko, A.G.
1996-01-01
This document contains the abstracts of the invited and contributed papers presented at 23 EPS conference on controlled fusion and plasma physics. The main contents are: tokamaks, stellarators; alternative magnetic confinement; plasma edge physics; plasma heating and current drive; plasma diagnostics; basic collisionless plasma physics; high intensity laser produced plasmas and inertial confinement; low-temperature plasmas
Industry roles in the Tokamak Physics Experiment
International Nuclear Information System (INIS)
Thomassen, K.I.
1994-01-01
The Tokamak Physics Experiment (TPX) is the first major fusion project opportunity in many years for US industry. Both the TPX management and the Department of Energy's Office of Fusion Energy are committed to creating industry roles that are integrated throughout the project and that appropriately use the capabilities they offer. To address industry roles in TPX it is first appropriate to describe the collaborative national approach taken for this program. The Director of the Princeton Plasma Physics Laboratory (PPPL) was asked by DOE to set up this national team structure, and the current senior management positions and delegated responsibilities reflect that approach. While reporting lines and delegated roles are clear in the organization chart for TPX, one way to view, it, different from that of the individuals responsible upward through this management structure for various elements of the project, is through institutional responsibilities to the senior management team. In this view the management team relies on several national laboratories, each using industry contracts for major sub-systems and components, to execute the project. These responsibilities for design and for contracting are listed, showing that all major contracts will come through three national laboratories, forming teams for their responsible activities
Modeling plasma/material interactions during a tokamak disruption
International Nuclear Information System (INIS)
Hassanein, A.; Konkashbaev, I.
1994-10-01
Disruptions in tokamak reactors are still of serious concern and present a potential obstacle for successful operation and reliable design. Erosion of plasma-facing materials due to thermal energy dump during a disruption can severely limit the lifetime of these components, therefore diminishing the economic feasibility of the reactor. A comprehensive disruption erosion model which takes into account the interplay of major physical processes during plasma-material interaction has been developed. The initial burst of energy delivered to facing-material surfaces from direct impact of plasma particles causes sudden ablation of these materials. As a result, a vapor cloud is formed in front of the incident plasma particles. Shortly thereafter, the plasma particles are stopped in the vapor cloud, heating and ionizing it. The energy transmitted to the material surfaces is then dominated by photon radiation. It is the dynamics and the evolution of this vapor cloud that finally determines the net erosion rate and, consequently, the component lifetime. The model integrates with sufficient detail and in a self-consistent way, material thermal evolution response, plasma-vapor interaction physics, vapor hydrodynamics, and radiation transport in order to realistically simulate the effects of a plasma disruption on plasma-facing components. Candidate materials such as beryllium and carbon have been analyzed. The dependence of the net erosion rate on disruption physics and various parameters was analyzed and is discussed
Overview of physics research on the TCV tokamak
Czech Academy of Sciences Publication Activity Database
Fasoli, A.; Alberti, S.; Amorim, P.; Angioni, C.; Asp, E.; Behn, R.; Bencze, A.; Berrino, J.; Blanchard, P.; Bortolon, A.; Brunner, S.; Camenen, Y.; Cirant, S.; Coda, S.; Curchod, L.; DeMeijere, K.; Duval, B. P.; Fable, E.; Fasel, D.; Felici, F.; Furno, I.; Garcia, O.E.; Giruzzi, G.; Gnesin, S.; Goodman, T.; Graves, J.; Gudozhnik, A.; Gulejova, B.; Henderson, M.; Hogge, J. Ph.; Horáček, Jan; Joye, B.; Karpushov, A.; Kim, S.-H.; Laqua, H.; Lister, J. B.; Llobet, X.; Madeira, T.; Marinoni, A.; Marki, J.; Martin, Y.; Maslov, M.; Medvedev, S.; Moret, J.-M.; Paley, J.; Pavlov, I.; Piffl, Vojtěch; Piras, F.; Pitts, R.A.; Pitzschke, A.; Pochelon, A.; Porte, L.; Reimerdes, H.; Rossel, J.; Sauter, O.; Scarabosio, A.; Schlatter, C.; Sushkov, A.; Testa, D.; Tonetti, G.; Tskhakaya, D.; Tran, M. Q.; Turco, F.; Turri, G.; Tye, R.; Udintsev, V.; Véres, G.; Villard, L.; Weisen, H.; Zhuchkova, A.; Zucca, C.
2009-01-01
Roč. 49, č. 10 (2009), s. 104005-104005 ISSN 0029-5515 Institutional research plan: CEZ:AV0Z20430508 Keywords : overview highlights * fusion research * tokamak TCV * self-generated current * H-mode physics * Electron internal transport barrier * electron cyclotron heating * electron cyclotron current drive physics * density peaking * MHDactivity * edge physics * reciprocating Mach probe * Pfirsch–Schlueter component. Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://stacks.iop.org/NF/49/104005
Cairns, R A
1985-01-01
This book is intended as an introduction to plasma physics at a level suitable for advanced undergraduates or beginning postgraduate students in physics, applied mathematics or astrophysics. The main prerequisite is a knowledge of electromagnetism and of the associated mathematics of vector calculus. SI units are used throughout. There is still a tendency amongst some plasma physics researchers to· cling to C.g.S. units, but it is the author's view that universal adoption of SI units, which have been the internationally agreed standard since 1960, is to be encouraged. After a short introductory chapter, the basic properties of a plasma con cerning particle orbits, fluid theory, Coulomb collisions and waves are set out in Chapters 2-5, with illustrations drawn from problems in nuclear fusion research and space physics. The emphasis is on the essential physics involved and (he theoretical and mathematical approach has been kept as simple and intuitive as possible. An attempt has been made to draw attention t...
Plasma residual poloidal rotation in TCABR tokamak
International Nuclear Information System (INIS)
Severo, J.H.F.; Nascimento, I.C.; Tsypin, V.S.; Galvao, R.M.O.
2003-01-01
This paper reports the first measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak for a collisional plasma (Pfirsch-Schluter regime), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5±1.0)·10 3 m/s at r ∼ 2/3a, (a - limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) · 10 3 m/s for the plasma core to a co-current value of (2.0 ± 1.0) · 10 3 m/s near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner. (author)
Stability of tearing modes in tokamak plasmas
International Nuclear Information System (INIS)
Hegna, C.C.; Callen, J.D.
1994-02-01
The stability properties of m ≥ 2 tearing instabilities in tokamak plasmas are analyzed. A boundary layer theory is used to find asymptotic solutions to the ideal external kink equation which are used to obtain a simple analytic expression for the tearing instability parameter Δ'. This calculation generalizes previous work on this topic by considering more general toroidal equilibria (however, toroidal coupling effects are ignored). Constructions of Δ' are obtained for plasmas with finite beta and for islands that have nonzero width. A simple heuristic estimate is given for the value of the saturated island width when the instability criterion is violated. A connection is made between the calculation of the asymptotic matching parameter in the finite beta and island width case to the nonlinear analog of the Glasser effect
Sensitivity of transient synchrotron radiation to tokamak plasma parameters
International Nuclear Information System (INIS)
Fisch, N.J.; Kritz, A.H.
1988-12-01
Synchrotron radiation from a hot plasma can inform on certain plasma parameters. The dependence on plasma parameters is particularly sensitive for the transient radiation response to a brief, deliberate, perturbation of hot plasma electrons. We investigate how such a radiation response can be used to diagnose a variety of plasma parameters in a tokamak. 18 refs., 13 figs
Technology and plasma-materials interaction processes of tokamak disruptions
International Nuclear Information System (INIS)
McGrath, R.T.; Kellman, A.G.
1992-01-01
A workshop on the technology and plasma-materials interaction processes of tokamak disruptions was held April 3, 1992 in Monterey, California, as a satellite meeting of the 10th International Conference on Plasma-Surface Interactions. The objective was to bring together researchers working on disruption measurements in operating tokamaks, those performing disruption simulation experiments using pulsed plasma gun, electron beam and laser systems, and computational physicists attempting to model the evolution and plasma-materials interaction processes of tokamak disruptions. This is a brief report on the workshop. 4 refs
Triangularity effects on the collisional diffusion for elliptic tokamak plasma
International Nuclear Information System (INIS)
Martin, P.; Castro, E.
2007-01-01
In this conference the effect of ellipticity and triangularity will be analyzed for axisymmetric tokamak in the collisional regime. Analytic forms for the magnetic field cross sections are taken from those derived recently by other authors [1,2]. Analytical results can be obtained in elliptic plasmas with triangularity by using an special system of tokamak coordinates recently published [3-5]. Our results show that triangularities smaller than 0.6, increases confinement for ellipticities in the range 1.2 to 2. This behavior happens for negative and positive triangularities; however this effect is stronger for positive than for negative triangularities. The maximum diffusion velocity is not obtained for zero triangularity, but for small negative triangularities. Ellipticity is also very important in confinement, but the effect of triangularity seems to be more important. High electric inductive field increases confinement, though this field is difficult to modify once the tokamak has been built. The analytic form of the current produced by this field is like that of a weak Ware pinch with an additional factor, which weakens the effect by an order of magnitude. The dependence of the triangularity effect with the Shafranov shift is also analyzed. References 1. - L. L. Lao, S. P. Hirshman, and R. M. Wieland, Phys. Fluids 24, 1431 (1981) 2. - G. O. Ludwig, Plasma Physics Controlled Fusion 37, 633 (1995) 3. - P. Martin, Phys. Plasmas 7, 2915 (2000) 4. - P. Martin, M. G. Haines and E. Castro, Phys. Plasmas 12, 082506 (2005) 5. - P. Martin, E. Castro and M. G. Haines, Phys. Plasmas 12, 102505 (2005)
Divertor design for the Tokamak Physics Experiment
International Nuclear Information System (INIS)
Hill, D.N.; Braams, B.
1994-05-01
In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4x L-mode), high beta (β N ≥ 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74 degrees from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m 2 with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities
Neoclassical transport of impurtities in tokamak plasmas
International Nuclear Information System (INIS)
Hirshman, S.P.; Sigmar, D.J.
1981-05-01
Tokamak plasmas are inherently comprised of multiple ion species. This is due to wall-bred impurities and, in future reactors, will result from fusion-born alpha particles. Relatively small concentrations of highly charged non-hydrogenic impurities can strongly influence plasma transport properties whenever n/sub I/e/sub I/ 2 /n/sub H/e 2 greater than or equal to (m/sub e//m/sub H/)/sup 1/2/. The determination of the complete neoclassical Onsager matrix for a toroidally confined multispecies plasma, which provides the linear relation between the surface averaged radial fluxes and the thermodynamic forces (i.e., gradients of density and temperature, and the parallel electric field), is reviewed. A closed set of one-dimensional moment equations is presented for the time evolution of thermodynamic and magnetic field quantities which results from collisional transport of the plasma and two dimensional motion of the magnetic flux surface geometry. The effects of neutral beam injection on the equilibrium and transport properties of a toroidal plasma are consistently included
Plasma residual rotation in the TCABR tokamak
International Nuclear Information System (INIS)
Severo, J.H.F.; Nascimento, I.C.; Tsypin, V.S.; Galvao, R.M.O.
2003-01-01
This paper reports the first results on the measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak, in the collisional regime (Pfirsch-Schluter), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5±1.0) x 10 3 m s -1 at r ∼ 2/3a,(a-limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) x 10 3 m s -1 , at the plasma core, to a co-current value of (2.0 ± 0.9) x 10 3 m s -1 near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner (1991 Phys. Fluids B 3 2050). (author)
Power supplies for plasma column control in COMPASS tokamak
Czech Academy of Sciences Publication Activity Database
Havlíček, Josef; Hauptmann, R.; Peroutka, Oldřich; Tadros, Momtaz; Hron, Martin; Janky, Filip; Vondráček, Petr; Cahyna, Pavel; Mikulín, Ondřej; Šesták, David; Junek, Pavel; Pánek, Radomír
2013-01-01
Roč. 88, 9-10 (2013), s. 1640-1645 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * Power supplies * Feedback control * Vertical displacement * Vertical kicks Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001543#
Extremely shaped plasmas to improve the Tokamak concept
Energy Technology Data Exchange (ETDEWEB)
Piras, F.
2011-04-15
Energy is essential for human existence and our future depends on plentiful and accessible sources of energy. The world population is fast growing and the average energy used per capita increases. One of the greatest challenges for human beings is that of meeting the growing demand for energy in a responsible, equitable and sustainable way. The possibility to obtain energy by ‘fusing’ light atoms addresses these needs. Nuclear fusion reactions are clean, safe and the amount of fuel present on Earth (hydrogen isotopes) is practically inexhaustible and well distributed. Nuclear fusion is a natural process that occurs in all active stars like our Sun. Since the first demonstration of a deuterium fusion reaction (Rutherford 1933), researchers worldwide have tried to replicate this process on Earth by building a thermonuclear fusion reactor. Nevertheless, the challenge posed by the construction of a nuclear fusion reactor is greater than the one presented earlier by the development of a fission reactor. During the IAEA Conference in Geneva in the early 1958, L.A. Artsimovich declared: ‘Plasma physics is very difficult. Worldwide collaboration is needed for progress’ and E.Teller, at the same conference: ‘Fusion technology is very complex. It is almost impossible to build a fusion reactor in this century’. They were right. The extremely high temperature and density necessary to fuse hydrogen isotopes makes it difficult indeed to create a successful fusion reactor. Even though the physics of the fusion reaction appears clear, we are still facing problems on the road towards building the ‘box’ that can efficiently confine the hot gas in the state of plasma. The best results so far have been obtained confining a plasma with strong magnetic fields in a toroidal configuration (‘tokamak’). The Centre de Recherches en Physique des Plasmas in Switzerland actively studies this promising configuration towards the development of a nuclear fusion reactor. The
Extremely shaped plasmas to improve the Tokamak concept
International Nuclear Information System (INIS)
Piras, F.
2011-04-01
Energy is essential for human existence and our future depends on plentiful and accessible sources of energy. The world population is fast growing and the average energy used per capita increases. One of the greatest challenges for human beings is that of meeting the growing demand for energy in a responsible, equitable and sustainable way. The possibility to obtain energy by ‘fusing’ light atoms addresses these needs. Nuclear fusion reactions are clean, safe and the amount of fuel present on Earth (hydrogen isotopes) is practically inexhaustible and well distributed. Nuclear fusion is a natural process that occurs in all active stars like our Sun. Since the first demonstration of a deuterium fusion reaction (Rutherford 1933), researchers worldwide have tried to replicate this process on Earth by building a thermonuclear fusion reactor. Nevertheless, the challenge posed by the construction of a nuclear fusion reactor is greater than the one presented earlier by the development of a fission reactor. During the IAEA Conference in Geneva in the early 1958, L.A. Artsimovich declared: ‘Plasma physics is very difficult. Worldwide collaboration is needed for progress’ and E.Teller, at the same conference: ‘Fusion technology is very complex. It is almost impossible to build a fusion reactor in this century’. They were right. The extremely high temperature and density necessary to fuse hydrogen isotopes makes it difficult indeed to create a successful fusion reactor. Even though the physics of the fusion reaction appears clear, we are still facing problems on the road towards building the ‘box’ that can efficiently confine the hot gas in the state of plasma. The best results so far have been obtained confining a plasma with strong magnetic fields in a toroidal configuration (‘tokamak’). The Centre de Recherches en Physique des Plasmas in Switzerland actively studies this promising configuration towards the development of a nuclear fusion reactor. The
Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bremond, S
1995-10-18
Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.
Plasma boundary experiments on DIII-D tokamak
International Nuclear Information System (INIS)
Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, P.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.K.; Owen, L.; Matthews, G.
1990-01-01
A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p T < or approx.10.7%, P(auxiliary)< or approx.20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions. (orig.)
Plasma boundary experiments on DIII-D tokamak
International Nuclear Information System (INIS)
Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, T.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.; Owen, L.; Matthews, G.
1990-06-01
A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p approx-lt 3 MA, β T approx-lt 10.7%, P(auxiliary) approx-lt 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma surface interactions. 41 refs., 8 figs., 1 tab
Trade studies of plasma elongation for next-step tokamaks
International Nuclear Information System (INIS)
Galambos, J.D.; Strickler, D.J.; Peng, Y.K.M.; Reid, R.L.
1988-09-01
The effect of elongation on minimum-cost devices is investigated for elongations ranging from 2 to 3. The analysis, carried out with the TETRA tokamak systems code, includes the effects of elongation on both physics (plasma beta limit) and engineering (poloidal field coil currents) issues. When ignition is required, the minimum cost occurs for elongations from 2.3 to 2.9, depending on the plasma energy confinement scaling used. Scalings that include favorable plasma current dependence and/or degradation with fusion power tend to have minimum cost at higher elongation (2.5-2.9); scalings that depend primarily on size result in lower elongation (/approximately/2.3) for minimum cost. For design concepts that include steady-state current-driven operation, minimum cost occurs at an elongation of 2.3. 12 refs., 13 figs
Integrated plasma control for high performance tokamaks
International Nuclear Information System (INIS)
Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.
2005-01-01
Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)
Multiscale coherent structures in tokamak plasma turbulence
International Nuclear Information System (INIS)
Xu, G. S.; Wan, B. N.; Zhang, W.; Yang, Q. W.; Wang, L.; Wen, Y. Z.
2006-01-01
A 12-tip poloidal probe array is used on the HT-7 superconducting tokamak [Li, Wan, and Mao, Plasma Phys. Controlled Fusion 42, 135 (2000)] to measure plasma turbulence in the edge region. Some statistical analysis techniques are used to characterize the turbulence structures. It is found that the plasma turbulence is composed of multiscale coherent structures, i.e., turbulent eddies and there is self-similarity in a relative short scale range. The presence of the self-similarity is found due to the structural similarity of these eddies between different scales. These turbulent eddies constitute the basic convection cells, so the self-similar range is just the dominant scale range relevant to transport. The experimental results also indicate that the plasma turbulence is dominated by low-frequency and long-wavelength fluctuation components and its dispersion relation shows typical electron-drift-wave characteristics. Some large-scale coherent structures intermittently burst out and exhibit a very long poloidal extent, even longer than 6 cm. It is found that these large-scale coherent structures are mainly contributed by the low-frequency and long-wavelength fluctuating components and their presence is responsible for the observations of long-range correlations, i.e., the correlation in the scale range much longer than the turbulence decorrelation scale. These experimental observations suggest that the coexistence of multiscale coherent structures results in the self-similar turbulent state
The major tokamak distruption in cylindrical plasma
International Nuclear Information System (INIS)
Choi, Jeong Sik; Choi, Eun Ha; Choi, Duk In
1986-01-01
The mechanism of the major disruption in tokamak plasma which involves the nonlinear interaction of tearing models is numerically studied in two and three dimensional formulations. In this study, it is found that in the two dimensional case with a flattened current density profile the magnetic islands of the m=2; n=1 mode do not saturate nonlinearly and but strongly interact with the limiter. Thus it is suggested that the helical perturbation of the m=2;n=1 mode plays the dominant role in the major disruption. We also show that the m=2;n=1 mode nonlinearly destablizes other tearing modes, especially the m=3;n=2 mode, from the nonlinear coupling of different helicities as also shown in other studies. The plasma extends across the plasma cross section, and the plasma core shifts inward along the major radius during the major disruption. The numerical result for the major disruption time measured using the nonlinear 3-D procedure for the initial value problem with PLT parameters is about 450 μsec which agrees reasonably well with the experimental value of 500 μsec. (Author)
Continuous, saturation, and discontinuous tokamak plasma vertical position control systems
Energy Technology Data Exchange (ETDEWEB)
Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)
2016-10-15
Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.
Continuous, saturation, and discontinuous tokamak plasma vertical position control systems
International Nuclear Information System (INIS)
Mitrishkin, Yuri V.; Pavlova, Evgeniia A.; Kuznetsov, Evgenii A.; Gaydamaka, Kirill I.
2016-01-01
Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.
Digital control of plasma position in Damavand tokamak
Energy Technology Data Exchange (ETDEWEB)
Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.
2002-03-01
Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)
A general comparison between tokamak and stellarator plasmas
Directory of Open Access Journals (Sweden)
Yuhong Xu
2016-07-01
Full Text Available This paper generally compares the essential features between tokamaks and stellarators, based on previous review work individually made by authors on several specific topics, such as theories, bulk plasma transport and edge divertor physics, along with some recent results. It aims at summarizing the main results and conclusions with regard to the advantages and disadvantages in these two types of magnetic fusion devices. The comparison includes basic magnetic configurations, magnetohydrodynamic (MHD instabilities, operational limits and disruptions, neoclassical and turbulent transport, confinement scaling and isotopic effects, plasma rotation, and edge and divertor physics. Finally, a concept of quasi-symmetric stellarators is briefly referred along with a comparison of future application for fusion reactors.
Control strategy for plasma equilibrium in a tokamak
International Nuclear Information System (INIS)
Miskell, R.V.
1975-08-01
Dynamic control of the plasma position within the torus of a TOKAMAK fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. (auth)
Neoclassical MHD descriptions of tokamak plasmas
International Nuclear Information System (INIS)
Callen, J.D.; Kim, Y.B.; Sundaram, A.K.
1988-01-01
Considerable progress has been made in extending neoclassical MHD theory and in exploring the linear instabilities, nonlinear behavior and turbulence models it implies for tokamak plasmas. The areas highlighted in this paper include: extension of the neoclassical MHD equations to include temperature-gradient and heat flow effects; the free energy and entropy evolution implied by this more complete description; a proper ballooning mode formalism analysis of the linear instabilities; a new rippling mode type instability; numerical simulation of the linear instabilities which exhibit a smooth transition from resistive ballooning modes at high collisionality to neoclassical MHD modes at low collisionality; numerical simulation of the nonlinear growth of a single helicity tearing mode; and a Direct-Interaction-Approximation model of neoclassical MHD turbulence and the anomalous transport it induces which substantially improves upon previous mixing length model estimates. 34 refs., 2 figs
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.
The COMPASS Tokamak Plasma Control Software Performance
Valcarcel, Daniel F.; Neto, André; Carvalho, Ivo S.; Carvalho, Bernardo B.; Fernandes, Horácio; Sousa, Jorge; Janky, Filip; Havlicek, Josef; Beno, Radek; Horacek, Jan; Hron, Martin; Panek, Radomir
2011-08-01
The COMPASS tokamak has began operation at the IPP Prague in December 2008. A new control system has been built using an ATCA-based real-time system developed at IST Lisbon. The control software is implemented on top of the MARTe real-time framework attaining control cycles as short as 50 μs, with a jitter of less than 1 μs. The controlled parameters, important for the plasma performance, are the plasma current, position of the plasma current center, boundary shape and horizontal and vertical velocities. These are divided in two control cycles: slow at 500 μs and fast at 50 μs. The project has two phases. First, the software implements a digital controller, similar to the analog one used during the COMPASS-D operation in Culham. In the slow cycle, the plasma current and position are measured and controlled with PID and feedforward controllers, respectively, the shaping magnetic field is preprogrammed. The vertical instability and horizontal equilibrium are controlled with the faster 50-μs cycle PID controllers. The second phase will implement a plasma-shape reconstruction algorithm and controller, aiming at optimized plasma performance. The system was designed to be as modular as possible by breaking the functional requirements of the control system into several independent and specialized modules. This splitting enabled tuning the execution of each system part and to use the modules in a variety of applications with different time constraints. This paper presents the design and overall performance of the COMPASS control software.
Physics design of the HL-2A tokamak
International Nuclear Information System (INIS)
Gao Qingdi; Li Fangzhu; Zhang Jinhua; Pan Yudong; Jiao Yiming
2005-10-01
An overview report for the physics design of the HL-2A tokamak is presented. By numerically analyzing the plasma shaping and the vertical instability due to plasma elongation, the requirements for the currents of poloidal magnetic field coils and the control system are put forward. Controlling the plasma profile by using NBI (neutral beam injection) and LHCD (lower hybrid current drive) is investigated, and the high performance modes of operation in HL-2A and modeled and designed. The magnetohydrodynamic instabilities in improved confinement configuration (RS configuration) are studied so as to point out the way of plasma control to perform stationary high performance discharges in HL-2A. In order to offer data for updating the HL-2A divertor, performances of the divertor plasma are simulated. (authors)
Self-organized ignition of a tokamak plasma
International Nuclear Information System (INIS)
Schoepf, K.
2007-01-01
The continuous progress in the attainment of plasma parameters required for establishing nuclear fusion in magnetically confined plasmas as well as the prospect of feasible steady-state operation has instigated the interest in the physics of burning plasmas [1]. Aside from the required plasma current drive, fusion energy production with tokamaks demands particular attention to confinement and fuelling regimes in order to maintain the plasma density n and temperature T at favourable values matching with specific requirements such as the triple product nτ E T, where τ E represents the plasma energy confinement time. The identification of state and parameter space regions capable of ignited fusion plasma operation is evidently crucial if significant energy gains are to be realized over longer periods. Examining the time-evolving state of tokamak fusion plasma in a parameter space spanned by the densities of plasma constituents and their temperatures has led to the formation of an ignition criterion [2] fundamentally different from the commonly used static patterns. The incorporation of non-stationary particle and energy balances into the analysis here, the application of a 'soft' Troyon beta limit [3], the consideration of actual fusion power deposition [4,5] and its effect of reducing τ E are seen to significantly influence the fusion burn dynamics and to shape the ignition conditions. The presented investigation refers to a somewhat upgraded (to achieve ignition) ITER-like tokamak plasma and uses volume averages of locally varying quantities and processes. The resulting ignition criterion accounts for the dynamic evolution of a reacting plasma controlled by heating and fuel feeding. Interestingly, also self-organized ignition can be observed: a fusion plasma possessing a density and temperature above a distinct separatrix in the considered parameter phase space is seen to evolve - without external heating and hence practically by itself - towards an ignited
International Nuclear Information System (INIS)
Barbosa, Luis Filipe F.P.W.; Bosco, Edson del.
1994-01-01
This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs
Kinetic theory of plasma adiabatic major radius compression in tokamaks
International Nuclear Information System (INIS)
Gorelenkova, M.V.; Gorelenkov, N.N.; Azizov, E.A.; Romannikov, A.N.; Herrmann, H.W.
1998-01-01
In order to understand the individual charged particle behavior as well as plasma macroparameters (temperature, density, etc.) during the adiabatic major radius compression (R-compression) in a tokamak, a kinetic approach is used. The perpendicular electric field from the Ohm close-quote s law at zero resistivity is made use of in order to describe particle motion during the R-compression. Expressions for both passing and trapped particle energy and pitch angle change are derived for a plasma with high aspect ratio and circular magnetic surfaces. The particle behavior near the passing trapped boundary during the compression is studied to simulate the compression-induced collisional losses of alpha particles. Qualitative agreement is obtained with the alphas loss measurements in deuterium-tritium (D-T) experiments in the Tokamak Fusion Test Reactor (TFTR) [World Survey of Activities in Controlled Fusion Research [Nucl. Fusion special supplement (1991)] (International Atomic Energy Agency, Vienna, 1991)]. The plasma macroparameters evolution at the R-compression is calculated by solving the gyroaveraged drift kinetic equation. copyright 1998 American Institute of Physics
The physics of tokamak start-up
International Nuclear Information System (INIS)
Mueller, D.
2013-01-01
Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current
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
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)
Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA
Indian Academy of Sciences (India)
The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is ﬁrst tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...
TFTR/JET INTOR workshop on plasma transport tokamaks
International Nuclear Information System (INIS)
Singer, C.E.
1985-01-01
This report summarizes the proceedings of a Workshop on transport models for prediction and analysis of tokamak plasma confinement. Summaries of papers on theory, predictive modeling, and data analysis are included
Characterizing electrostatic turbulence in tokamak plasmas with high MHD activity
Energy Technology Data Exchange (ETDEWEB)
Guimaraes-Filho, Z O; Santos Lima, G Z dos; Caldas, I L; Nascimento, I C; Kuznetsov, Yu K [Instituto de Fisica, Universidade de Sao Paulo, Caixa Postal 66316, 05315-970, Sao Paulo, SP (Brazil); Viana, R L, E-mail: viana@fisica.ufpr.b [Departamento de Fisica, Universidade Federal do Parana, Caixa Postal 19044, 81531-990, Curitiba, PR (Brazil)
2010-09-01
One of the challenges in obtaining long lasting magnetic confinement of fusion plasmas in tokamaks is to control electrostatic turbulence near the vessel wall. A necessary step towards achieving this goal is to characterize the turbulence level and so as to quantify its effect on the transport of energy and particles of the plasma. In this paper we present experimental results on the characterization of electrostatic turbulence in Tokamak Chauffage Alfven Bresilien (TCABR), operating in the Institute of Physics of University of Sao Paulo, Brazil. In particular, we investigate the effect of certain magnetic field fluctuations, due to magnetohydrodynamical (MHD) instabilities activity, on the spectral properties of electrostatic turbulence at plasma edge. In some TCABR discharges we observe that this MHD activity may increase spontaneously, following changes in the edge safety factor, or after changes in the radial electric field achieved by electrode biasing. During the high MHD activity, the magnetic oscillations and the plasma edge electrostatic turbulence present several common linear spectral features with a noticeable dominant peak in the same frequency. In this article, dynamical analyses were applied to find other alterations on turbulence characteristics due to the MHD activity and turbulence enhancement. A recurrence quantification analysis shows that the turbulence determinism radial profile is substantially changed, becoming more radially uniform, during the high MHD activity. Moreover, the bicoherence spectra of these two kinds of fluctuations are similar and present high bicoherence levels associated with the MHD frequency. In contrast with the bicoherence spectral changes, that are radially localized at the plasma edge, the turbulence recurrence is broadly altered at the plasma edge and the scrape-off layer.
A control approach for plasma density in tokamak machines
Energy Technology Data Exchange (ETDEWEB)
Boncagni, Luca, E-mail: luca.boncagni@enea.it [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Pucci, Daniele; Piesco, F.; Zarfati, Emanuele [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy); Mazzitelli, G. [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Monaco, S. [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy)
2013-10-15
Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1].
Experimental observations related to the thermodynamic properties of tokamak plasmas
International Nuclear Information System (INIS)
Sozzi, C.; Minardi, E.; Lazzaro, E.; Cirant, S.; Mantica, P.; Esposito, B.; Marinucci, M.; Romanelli, M.; Imbeaux, F.
2005-01-01
The coarse-grained tokamak plasma description derived from the magnetic entropy concept presents appealing features as it involves a simple mathematics and it identifies a limited set of characteristic parameters of the macroscopic equilibrium. In this paper a comprehensive review of the work done in order to check the reliability of the Stationary Magnetic Entropy predictions against experimental data collected from different tokamaks, plasma regimes and heating methods is reported. (author)
International Nuclear Information System (INIS)
1988-05-01
The Steady-state Tokamak (STE) Experiment is a proposed superconducting-coil, hydrogen-plasma tokamak device intended to address the integrated non-nuclear issues of steady state, high-power tokamak physics and technology. Such a facility has been called for in the US program plan for the mid 1990's, and will play a unique role in the world-wide fusion effort. Information from STE on steady-state current drive, plasma control, and high power technology will contribute significantly to the operating capabilities of future steady-state devices. This paper reviews preliminary designs and expected technological contributions to the US and world fusion reactor research from each of the above mentioned reactor systems. This document is intended as a proposal and feasibility discussion and does not include exhaustive technical reviews. 12 figs., 3 tabs
Boriev, I. A.
2018-03-01
An analysis of the problem of so-called “abnormal” fast transfer of electrons in tokamak plasma, which turned out much faster than the result of accepted calculation, is given. Such transfer of hot electrons leads to unexpectedly fast destruction of the inner tokamak wall with ejection of its matter in plasma volume, what violates a condition of plasma confinement for controlled thermonuclear fusion. It is shown, taking into account real physics of electron drift in the gas (plasma) and using the conservation law for momentum of electron transfer (drift), that the drift velocity of elastically scattered electrons should be significantly greater than that of accepted calculation. The reason is that the relaxation time of the momentum of electron transfer, to which the electron drift velocity is proportional, is significantly greater (from 16 up to 4 times) than the electron free path time. Therefore, generally accepted replacement of the relaxation time, which is unknown a priori, by the electron free path time, leads to significant (16 times for thermal electrons) underestimation of electron drift velocity (mobility). This result means, that transfer of elastically (and isotropically) scattered electrons in the gas phase should be so fast, and corresponds to multiplying coefficient (16), introduced by D. Bohm to explain the observed by him “abnormal” fast diffusion of electrons.
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.
Plasma Confinement in the UCLA Electric Tokamak.
Taylor, Robert J.
2001-10-01
The main goal of the newly constructed large Electric Tokamak (R = 5 m, a = 1 m, BT 8 x 10^12 cm-3 when there is no MHD activity. The electron temperature, derived from the plasma conductivity is > 250 eV with a central electron energy confinement time > 350 msec in ohmic conditions. The sawteeth period is 50 msec. Edge plasma rotation is induced by plasma biasing via electron injection in an analogous manner to that seen in CCT(R.J. Taylor, M.L. Brown, B.D. Fried, H. Grote, J.R. Liberati, G.J. Morales, P. Pribyl, D. Darrow, and M. Ono. Phys. Rev Lett. 63 2365 1989.) and the neoclassical bifurcation is close to that described by Shaing et al(K.C. Shaing and E.C. Crume, Phys. Rev. Lett. 63 2369 (1989).). In the ohmic phase the confinement tends to be MHD limited. The ICRF heating eliminates the MHD disturbances. Under second harmonic heating conditions, we observe an internal confinement peaking characterized by doubling of the core density and a corresponding increase in the central electron temperature. Charge exchange data, Doppler data in visible H-alpha light, and EC radiation all indicate that ICRF heating works much better than expected. The major effort is focused on increasing the power input and controlling the resulting equilibrium. This task appears to be easy since our current pulses are approaching the 3 second mark without RF heating or current drive. Our initial experience with current profile control, needed for high beta plasma equilibrium, will be also discussed.
Long pulse neutral beam system for the Tokamak Physics Experiment
International Nuclear Information System (INIS)
Grisham, L.R.; Bowen, O.N.; Dahlgren, F.; Edwards, J.W.; Kamperschroer, J.; Newman, R.; O'Connor, T.; Ramakrishnan, S.; Rossi, G.; Stevenson, T.; Halle, A. von; Wright, K.E.
1995-01-01
The Tokamak Physics Experiment (TPX) is planned as a long-pulse or steady-state machine to serve as a successor to the Tokamak Fusion Test Reactor (TFTR). The neutral beam component of the heating and current drive systems will be provided by a TFTR beamline modified to allow operation for pulse lengths of 1000s. This paper presents a brief overview of the conceptual design which has been carried out to determine the changes to the beamline and power supply components that will be required to extend the pulse length from its present limitation of 1s at full power. The modified system, like the present one, will be capable of injecting about 8MW of power as neutral deuterium. The initial operation will be with a single beamline oriented co-directional to the plasma current, but the TPX system design is capable of accommodating an additional co-directional beamline and a counter-directional beamline. ((orig.))
Trapping of gun-injected plasma by a tokamak
International Nuclear Information System (INIS)
Leonard, A.W.; Dexter, R.N.; Sprott, J.C.
1986-01-01
It is shown that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. Gun injection raises the line-averaged density and peaks the density profile. Trapping of the gun-injected plasma is explainable in terms of a depolarization current mechanism
Measurements of plasma position in TJ-I Tokamak
International Nuclear Information System (INIS)
Qin, J.; Ascasibar, E.; Navarro, A.P.; Ochando, M.A.; Pastor, I.; Pedrosa, M.A.; Rodriguez, L.; Sanchez, J.; Team, TJ-I.
1994-01-01
This report presents the experimental measurements of plasma position in TJ-I tokamak by using small magnetic probes. The basis of method has been described in our previous work (1) in which the plasma current is considered as a filament current. The observed relations between the disruptive instabilities and plasma displacements are also show here. (Author) 7 refs
International Conference on Plasma Physics ICPP 1994. Proceedings
International Nuclear Information System (INIS)
Sakanaka, P.H.; Tendler, M.
1995-01-01
These proceedings represent the papers presented at the 1994 International Conference on Plasma Physics held in Foz do Iguacu, Brazil. The scope of the conference was broad and covered all aspects of plasma physics. Some of the topics discussed include space and astrophysical plasmas,fusion plasmas, small and large Tokamak plasmas, non-Tokamak plasmas, inertial confinement fusion plasmas, plasma based neutron sources and plasma applications. There are 60 papers in these proceedings and out of these, 35 have been abstracted for the Energy Science and Technology database
Effect of neutral atoms on tokamak edge plasmas
International Nuclear Information System (INIS)
Fueloep, T.; Catto, Peter J.; Helander, P.
2001-01-01
Neutral atoms can significantly influence the physics of tokamak edge plasmas, e.g., by affecting the radial electric field and plasma flow there, which may, in turn, be important for plasma confinement. Earlier work [Fueloep et al., Phys. Plasmas 5, 3969 (1998)], assuming short mean-free path neutrals and Pfirsch-Schlueter ions, has shown that the ion-neutral coupling through charge-exchange affects the neoclassical flow velocity significantly. However, the mean-free path of the neutrals is not always small in comparison with the radial scale length of densities and temperatures in the edge pedestal. It is therefore desirable to determine what happens in the limit when the neutral mean-free path is comparable with the scale length. In the present work a self-similar solution for the neutral distribution function allowing for strong temperature and density variation is used, following the analysis of Helander and Krasheninnikov [Phys. Plasmas 3, 226 (1995)]. The self-similar solution is possible if the ratio of the mean-free path to the temperature and density scale length is constant throughout the edge plasma. The resulting neutral distribution function is used to investigate the neutral effects on the ion flow and electrostatic potential as this ratio varies from much less than one to order unity
Bifurcated states of a rotating tokamak plasma in the presence of a static error-field
International Nuclear Information System (INIS)
Fitzpatrick, R.
1998-01-01
The bifurcated states of a rotating tokamak plasma in the presence of a static, resonant, error-field are strongly analogous to the bifurcated states of a conventional induction motor. The two plasma states are the open-quotes unreconnectedclose quotes state, in which the plasma rotates and error-field-driven magnetic reconnection is suppressed, and the open-quotes fully reconnectedclose quotes state, in which the plasma rotation at the rational surface is arrested and driven magnetic reconnection proceeds without hindrance. The response regime of a rotating tokamak plasma in the vicinity of the rational surface to a static, resonant, error-field is determined by three parameters: the normalized plasma viscosity, P, the normalized plasma rotation, Q 0 , and the normalized plasma resistivity, R. There are 11 distinguishable response regimes. The extents of these regimes are calculated in P endash Q 0 endash R space. In addition, an expression for the critical error-field amplitude required to trigger a bifurcation from the open-quotes unreconnectedclose quotes to the open-quotes fully reconnectedclose quotes state is obtained in each regime. The appropriate response regime for low-density, ohmically heated, tokamak plasmas is found to be the nonlinear constant-ψ regime for small tokamaks, and the linear constant-ψ regime for large tokamaks. The critical error-field amplitude required to trigger error-field-driven magnetic reconnection in such plasmas is a rapidly decreasing function of machine size, indicating that particular care may be needed to be taken to reduce resonant error-fields in a reactor-sized tokamak. copyright 1998 American Institute of Physics
Control of plasma position in the CASTOR tokamak
International Nuclear Information System (INIS)
Valovic, M.
1988-11-01
A simple servo-system designed for plasma position control in the CASTOR tokamak is described. Both radial and vertical plasma displacements were minimized using two servo-loops consisting of detection coils, a conventional electric controller and an amplifier operated as an unipolar voltage-controlled current source. To ensure the optimum conditions in the start-up phase of the discharge, currents in the servo-systems were externally preprogrammed. The prescribed plasma position was maintained with the accuracy of 3 mm. The feedback control improves plasma parameters, e.g. it removes the positional disruption at the end of the tokamak discharge. (J.U.). 4 figs., 3 refs
Magnetic diagnostic plasma position in the TCA/BR tokamak
International Nuclear Information System (INIS)
Galvao, R.M.O.; Kuznetsov, Yu.K.; Nascimento, I.C.
1996-01-01
The cross-section of the plasma column is TCA/BR has a nearly circular plasma shape. This allows implementation of simplified methods of magnetic diagnostics. Although these methods were in may tokamaks and are well described, their accuracies are not clearly defined because the very simplified theoretical model of plasma equilibrium on which they are based differs from the real conditions in tokamaks like TCA/BR. In this paper we present the methods of plasma position diagnostics in TCA/BR from external magnetic measurements with an error analysis. (author). 4 refs., 3 figs
Physics of transport in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Garbet, X [Assoc. EURATOM-CEA, CEA/DSM/DRFC CEA-Cadarache, 13108 Saint Paul lez Durance (France); Mantica, P [Istituto di Fisica del Plasma EURATOM-ENEA/CNR, via Cozzi 53, 20125 Milan (Italy); Angioni, C [MPI fuer Plasmaphysik, EURATOM-Assoz., D-8046 Garching (Germany)] [and others
2004-12-01
This paper is an overview of recent results relating to turbulent particle and heat transport, and to the triggering of internal transport barriers (ITBs). The dependence of the turbulent particle pinch velocity on plasma parameters has been clarified and compared with experiment. Magnetic shear and collisionality are found to play a central role. Analysis of heat transport has made progress along two directions: dimensionless scaling laws, which are found to agree with the prediction for electrostatic turbulence, and analysis of modulation experiments, which provide a stringent test of transport models. Finally the formation of ITBs has been addressed by analysing electron transport barriers. It is confirmed that negative magnetic shear, combined with the Shafranov shift, is a robust stabilizing mechanism. However, some well established features of internal barriers are not explained by theory.
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
DAMAVAND - An Iranian tokamak with a highly elongated plasma cross-section
International Nuclear Information System (INIS)
Amrollahi, R.
1997-01-01
The ''DAMAVAND'' facility is an Iranian Tokamak with a highly elongated plasma cross-section and with a poloidal divertor. This Tokamak has the advantage to allow the plasma physics research under the conditions similar to those of ITER magnetic configuration. For example, the opportunity to reproduce partially the plasma disruptions without sacrificing the studies of: equilibrium, stability and control over the elongated plasma cross-section; processes in the near-wall plasma; auxiliary heating systems, etc. The range of plasma parameters, the configuration of ''DAMAVAND'' magnetic coils and passive loops, and their location within the vacuum chamber allow the creation of the plasma at the center of the vacuum chamber and the production of two poloidal volumes (upper and lower) for the divertor. (author)
Orbit effects on impurity transport in a rotating tokamak plasma
International Nuclear Information System (INIS)
Wong, K.L.; Cheng, C.Z.
1988-05-01
Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster with a higher bounce frequency, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle orbits near the surface of a rotating tokamak are also analyzed. Orbit effects indicate that more impurities can penetrate into a plasma rotating with counter-beam injection. Particle simulation is carried out with realistic experimental parameters and the results are in qualitative agreement with some experimental observations in the Tokamak Fusion Test Reactor (TFTR). 19 refs., 15 figs
HT-7U superconducting tokamak: Physics design, engineering progress and schedule
International Nuclear Information System (INIS)
Wan Yuanxi
2002-01-01
The superconducting tokamak research program begun in China in ASIPP since 1994. The program is included in existent superconducting tokamak HT-7 and the next new superconducting tokamak HT-7U which is one of national key research projects in China. With the elongation cross-section, divertor and higher plasma parameter the main objectives of HT-7U are widely investigation both of the physics and technology for steady state advanced tokamak as well as the investigation of power and particle handle under steady-state operation condition. The physics and engineering design have been completed and significant progresses on R and D and fabrication have been achieved. HT-7U will begin assembly at 2003 and possible to get first plasma around 2004. (author)
Development of Tokamak experiment technology - Study of ICRF coupling in the KAIST tokamak plasma
Energy Technology Data Exchange (ETDEWEB)
Choi, Duk In; Chang, Hang Young; Lee, Soon Chil; Kwon, Gi Chung; Seo, Sung Hun; Jeon, Sang Jin; Heo, Sung Hee; Heo, Eun Gi; Lee, Dae Hang; Lee, Chan Hee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
1995-08-01
Research objectives are to design and fabricate antenna, measure the property of absorption transmitted to the plasma, and research the physical phenomena about the ICRF coupling. Main heating method is ohmic heating at the KAIST tokamak. So, the plasma current produced is more than 30 kA and, the loop voltage of the plasma is 2 {approx} 3V. The power of the plasma by ohmic heating is about 100 kW. Because the toroidal field is 5 {approx} 8 kG, it is needed RF system with more than 100 kW in 7 {approx} 15 MHz. In the first year a RF amplifier with 1 kW in 300 khz {approx} 35 MHz was bought. The manufacture of ICRF system will start from next years. In the research on antenna, we study the method how to measure electric field emitted from antenna using piezo elements. Experimentally, we obtain the results that the signal of piezo element is proportional to the square of electric field. In the next year, we will research the type of antenna subsequently. 28 refs., 3 tabs., 18 figs. (author)
Edge Plasma Response to Non-Axisymmetric Fields in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Ferraro, N. M.; Lao, L. L.; Buttery, R. J.; Evans, T. E.; Snyder, P. B.; Wade, M.R., E-mail: ferraro@fusion.gat.com [General Atomics, San Diego (United States); Moyer, R. A.; Orlov, D. M. [University of California San Diego, La Jolla (United States); Lanctot, M. J. [Lawrence Livermore National Laboratory, Livermore (United States)
2012-09-15
Full text: The application of non-axisymmetric fields is found to have significant effects on the transport and stability of H-mode tokamak plasmas. These effects include dramatic changes in rotation and particle transport, and may lead to the partial or complete suppression of edge-localized modes (ELMs) under some circumstances. The physical mechanism underlying these effects is presently not well understood, in large part because the response of the plasma to non- axisymmetric fields is significant and complex. Here, recent advances in modeling the plasma response to non-axisymmetric fields are discussed. Calculations using a resistive two-fluid model in diverted toroidal geometry confirm the special role of the perpendicular electron velocity in suppressing the formation of islands in the plasma. The possibility that islands form near the top of the pedestal, where the zero-crossing of the perpendicular electron velocity may coincide with a mode-rational surface, is explored, and the implications for ELM suppression are discussed. Modeling results are compared with empirical data. It is shown that numerical modeling is successful in reproducing some experimentally observed effects of applied non-axisymmetric fields on the edge temperature and density profiles. The numerical model self-consistently includes the plasma, separatrix, and scrape-off layer. Rotation and diamagnetic effects are also included self-consistently. Solutions are calculated using the M3D-C1 extended-MHD code. (and others)
Physics of transport in Tokamaks
International Nuclear Information System (INIS)
Garbet, X.; Asp, E.; Bourdelle, C.; Hoang, T.; Imbeaux, F.; Joffrin, E.; Litaudon, X.; Angioni, C.; Manini, A.; Peeter, A.; Ryter, F.; Baranov, Y.; Cordey, G.; McDonald, D.C.; Parail, V.; Valovic, M.; Thyagaraja, A.; Voitsekhovitch, I.; Budny, R.; Kirneva, N.; Hogeweij, D.; Nordman, H.; Weilland, J.; Tala, T.; Weisen, H.; Zabolotsky, A.
2004-01-01
This paper is an overview of recent results related to turbulent particle and heat transport, and triggering of Internal Transport Barriers. Particle transport is characterised by a pinch velocity that is found to be larger than the value predicted by theory of collisional transport (Ware pinch) in L-mode plasmas. Also it increases with magnetic shear. In H-mode, density peaking decreases with collisionality. Pinch velocity reaches the Ware value for large collision frequencies. Heat transport has made progress along two directions: dimensionless scaling laws and analysis of modulation experiments. Dimensionless scaling law of thermal confinement agrees with the prediction for electrostatic turbulence. Dependence with collisionality remains to be understood. Heat modulation experiments have been investigated in several devices using a critical gradient model, micro-stability analysis and predictive modelling. Thresholds and stiffness are correctly reproduced by stability analysis and modelling with Weiland model. Analysis with a critical gradient model leads to a large variability of stiffness. Finally the question of triggering Internal Transport Barriers has been addressed by analysing electron transport barriers. It is confirmed that negative magnetic shear combined with Shafranov shift is a robust stabilizing mechanism. However, some well established features of internal barriers are hardly explained by theory, in particular the role of low order rational values of the minimum safety factor, and the existence of multiple barriers. (authors)
Diagnosing transient plasma status: from solar atmosphere to tokamak divertor
International Nuclear Information System (INIS)
Giunta, A.S.; Henderson, S.; O'Mullane, M.; Summers, H.P.; Harrison, J.; Doyle, J.G.
2016-01-01
This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.
Magnetic analysis of tokamak plasma with approximate MHD equilibrium solution
International Nuclear Information System (INIS)
Moriyama, Shin-ichi; Hiraki, Naoji
1993-01-01
A magnetic analysis method for determining equilibrium configuration parameters (plasma shape, poloidal beta and internal inductance) on a non-circular tokamak is described. The feature is to utilize an approximate MHD equilibrium solution which explicitly relates the configuration parameters with the magnetic fields picked up by magnetic sensors. So this method is suitable for the real-time analysis performed during a tokamak discharge. A least-squares fitting procedure is added to the analytical algorithm in order to reduce the errors in the magnetic analysis. The validity is investigated through the numerical calculation for a tokamak equilibrium model. (author)
Physics objectives of PI3 spherical tokamak program
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.
Behaviour of metallic droplets in a tokamak plasma
International Nuclear Information System (INIS)
Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.
1989-01-01
Micrometre sized tantalum droplets were injected into a tokamak plasma by a controllable arcing gun located behind the wall. The trajectories of the ablating particles were photographed by a high speed camera. Various possible mechanisms which may explain the observed curvature of the particle paths are discussed. The migration of the ablated material in the tokamak was studied by post-mortem analysis of collector probes and limiters. (author). Letter-to-the-editor. 12 refs, 9 figs
Tokamak physics experiment: Diagnostic windows study
International Nuclear Information System (INIS)
Merrigan, M.; Wurden, G.A.
1995-11-01
We detail the study of diagnostic windows and window thermal stress remediation in the long-pulse, high-power Tokamak Physics Experiment (TPX) operation. The operating environment of the TPX diagnostic windows is reviewed, thermal loads on the windows estimated, and cooling requirements for the windows considered. Applicable window-cooling technology from other fields is reviewed and its application to the TPX windows considered. Methods for TPX window thermal conditioning are recommended, with some discussion of potential implementation problems provided. Recommendations for further research and development work to ensure performance of windows in the TPX system are presented
A Midsize Tokamak As Fast Track To Burning Plasmas
International Nuclear Information System (INIS)
Mazzucato, E.
2010-01-01
This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain ((ge) 10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.
Scaling for scrape-off layer plasma in tokamak
International Nuclear Information System (INIS)
Shimomura, Yasuo; Maeda, Hikosuke; Kimura, Haruyuki; Azumi, Masashi; Odajima, Kazuo
1977-12-01
Scaling for a scrape-off layer plasma in a tokamak is obtained by using DIVA (JFT-2a). The scaling gives the average electron temperature, the width and the mean electron density of the scrape-off layer. The temperature at the edge will be high in a future large tokamak with a small energy-loss by charge-exchange and radiation. The scrape-off layer plasma can easily shield the impurity influx from the wall. The fuel, however, can easily penetrate into the main plasma. (auth.)
Energy Technology Data Exchange (ETDEWEB)
Fraboulet, D.
1996-09-17
Detection of {alpha}(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, {alpha} particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. {alpha} particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central {alpha} density in a reactor. (author). 166 refs.
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.
Feedback control of plasma position in the HL-1 tokamak
International Nuclear Information System (INIS)
Yuan Baoshan; Jiao Boliang; Yang Kailing
1991-01-01
In the HL-1 tokamak with a thick copper shell, the control of plasma position is successfully performed by a feedback-feedforward system with dual mode regulator and the equilibrium field coils outside the shell. The plasma position can be controlled within ±2 mm in both vertical and horizontal directions under the condition that the iron core of transformer is not saturated
Negative edge plasma currents in the SINP tokamak
Indian Academy of Sciences (India)
RAE is the maximum runaway energy emitted during a burst period of tdur. HXR. There being no plasma control feedback system in the SINP tokamak, the dynamics of the plasma equilibrium is time-dependent and the column shift is now made by the discharge dynamics itself. We measured DRAE for the two discharges ...
Plasma fluctuation measurements in tokamaks using beam-plasma interactions
International Nuclear Information System (INIS)
Fonck, R.J.; Duperrex, P.A.; Paul, S.F.
1990-01-01
High-frequency observations of light emitted from the interactions between plasma ions and injected neutral beam atoms allow the measurement of moderate-wavelength fluctuations in plasma and impurity ion densities. To detect turbulence in the local plasma ion density, the collisionally excited fluorescence from a neutral beam is measured either separately at several spatial points or with a multichannel imaging detector. Similarly, the role of impurity ion density fluctuations is measured using charge exchange recombination excited transitions emitted by the ion species of interest. This technique can access the relatively unexplored region of long-wavelength plasma turbulence with k perpendicular ρ i much-lt 1, and hence complements measurements from scattering experiments. Optimization of neutral beam geometry and optical sightlines can result in very good localization and resolution (Δx≤1 cm) in the hot plasma core region. The detectable fluctuation level is determined by photon statistics, atomic excitation processes, and beam stability, but can be as low as 0.2% in a 100 kHz bandwidth over the 0--1 MHz frequency range. The choices of beam species (e.g., H 0 , He 0 , etc.), observed transition (e.g., H α , L α , He I singlet or triplet transitions, C VI Δn=1, etc.) are dictated by experiment-specific factors such as optical access, flexibility of beam operation, plasma conditions, and detailed experimental goals. Initial tests on the PBX-M tokamak using the H α emissions from a heating neutral beam show low-frequency turbulence in the edge plasma region
Simulations of Neon Pellets for Plasma Disruption Mitigation in Tokamaks
Bosviel, Nicolas; Samulyak, Roman; Parks, Paul
2017-10-01
Numerical studies of the ablation of neon pellets in tokamaks in the plasma disruption mitigation parameter space have been performed using a time-dependent pellet ablation model based on the front tracking code FronTier-MHD. The main features of the model include the explicit tracking of the solid pellet/ablated gas interface, a self-consistent evolving potential distribution in the ablation cloud, JxB forces, atomic processes, and an improved electrical conductivity model. The equation of state model accounts for atomic processes in the ablation cloud as well as deviations from the ideal gas law in the dense, cold layers of neon gas near the pellet surface. Simulations predict processes in the ablation cloud and pellet ablation rates and address the sensitivity of pellet ablation processes to details of physics models, in particular the equation of state.
The COMPASS Tokamak Plasma Control Software Performance
Czech Academy of Sciences Publication Activity Database
Valcárcel, D.F.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Janky, F.; Havlíček, Josef; Beňo, R.; Horáček, Jan; Hron, Martin; Pánek, Radomír
2011-01-01
Roč. 58, č. 4 (2011), s. 1490-1496 ISSN 0018-9499. [Real Time Conference, RT10/17th./. Lisboa, 24.05.2010-28.05.2010] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-Time * ATCA * Data Acquisition * Plasma Control Software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.447, year: 2011 http://dx.doi.org/10.1109/TNS.2011.2143726
Study of Globus-M Tokamak Poloidal System and Plasma Position Control
Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.
2017-12-01
In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.
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
On thermionic emission from plasma-facing components in tokamak-relevant conditions.
Czech Academy of Sciences Publication Activity Database
Komm, Michael; Ratynskaia, S.; Tolias, P.; Cavalier, Jordan; Dejarnac, Renaud; Gunn, J. P.; Podolník, Aleš
2017-01-01
Roč. 59, č. 9 (2017), č. článku 094002. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : thermionic * PIC * tungsten * tokamak * thermionic emission * plasma facing components * particle-in-cell 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/10.1088/1361-6587/aa78c4/pdf
Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe
Czech Academy of Sciences Publication Activity Database
Walkden, N.R.; Adámek, Jiří; Allan, S.; Dudson, B.D.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Komm, Michael
2015-01-01
Roč. 86, č. 2 (2015), č. článku 023510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * ball pen probe Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.336, year: 2015 http://dx.doi.org/10.1063/1.4908572
Two-body similarity and its violation in tokamak edge plasmas
International Nuclear Information System (INIS)
Catto, P.J.; Knoll, D.A.; Krasheninnikov, S.I.
1996-01-01
Scaling laws found under the assumption that two-body collisions dominate can be effectively used to benchmark complex multi-dimensional codes dedicated to investigating tokamak edge plasmas. The applicability of such scaling laws to the interpretation of experimental data, however, is found to be restricted to the relatively low plasma densities ( 19 m -3 ) at which multistep processes, which break the two-body collision approximation, are unimportant. copyright 1996 American Institute of Physics
Reactor aspects of counterstreaming-ion tokamak plasmas
International Nuclear Information System (INIS)
Jassby, D.L.
1975-06-01
Toroidal DT plasmas in which the D and T ions make up two distinct, quasi-thermal velocity distributions, oppositely displaced in velocity along the magnetic axis, are discussed. Such counterstreaming distributions can be set up by introducing all ions by tangential injection of neutral beams, and by removing ions from the plasma shortly after they have decelerated to an energy approximate to or less than 2T/sub e/ by Coulomb drag on the plasma electrons. A simple physical model for counterstreaming-ion operation is postulated, which allows one to deduce the ion velocity distributions and required energy and particle confinement times that are in good agreement with the results of previous Fokker-Planck calculations. The variations of fusion reactivity, power gain, and power density with injection energy and electron temperature are presented. The practical problems of implementing counter-streaming operation in a tokamak, such as charge-exchange losses, the prompt removal of cold ions, and the effect of impurities are discussed. (U.S.)
Relaxation of plasma potential and poloidal flows in the boundary of tokamak plasmas
International Nuclear Information System (INIS)
Hron, M.; Duran, I.; Stoeckel, J.; Hidalgo, C.; Gunn, J.
2003-01-01
The relaxation times of plasma parameters after a sudden change of electrode voltage have been measured in the plasma boundary during polarization experiments on the CASTOR tokamak (R = 0.4 m, a = 75 mm, B t = 1 T, I p ∼ 9 kA, q a ∼ 10). The time evolution of the floating potential after the biasing voltage switch-off can be well fitted by an exponential decay with characteristic time in the range of 10 - 20 μs. The poloidal flow shows a transient behaviour with a time scale of about 10 - 30 μs. These time scales are smaller than the expected damping time based on neoclassical parallel viscosity (which is in the range of 100 νs) and atomic physics via charge exchange (in the range of 100 - 1000 νs). But, they are larger than the correlation time of plasma turbulence (about 5 μs). These findings suggest that anomalous damping rate mechanisms for radial electric fields and poloidal flows may play a role in the boundary of tokamak plasmas. (authors)
Internal transport barrier in tokamak and helical plasmas
Ida, K.; Fujita, T.
2018-03-01
The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, R/{L}T, the location, {ρ }{ITB}, and the width, W/a, and can be expressed by ‘weak’ ITB (small R/{L}T) or ‘strong’ (large R/{L}T), ‘small’ ITB (small {ρ }{ITB}) or ‘large’ ITB (large {ρ }{ITB}), and ‘narrow’ (small W/a) or ‘wide’ (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the
The superconducting magnet system for the Tokamak Physics Experiment
International Nuclear Information System (INIS)
Lang, D.D.; Bulmer, R.J.; Chaplin, M.R.; O'Connor, T.G.; Slack, D.S.; Wong, R.L.; Zbasnik, J.P.; Schultz, J.H.; Diatchenko, N.; Montgomery, D.B.
1994-01-01
The superconducting magnet system for the Tokamak Physics eXperiment (TPX) will be the first all superconducting magnet system for a Tokamak, where the poloidal field coils, in addition to the toroidal field coils are superconducting. The magnet system is designed to operate in a steady state mode, and to initiate the plasma discharge ohmically. The toroidal field system provides a peak field of 4.0 Tesla on the plasma axis at a plasma major radius of 2.25 m. The peak field on the niobium 3-tin, cable-in-conduit (CIC) conductor is 8.4 Tesla for the 16 toroidal field coils. The toroidal field coils must absorb approximately 5 kW due to nuclear heating, eddy currents, and other sources. The poloidal field system provides a total of 18 volt seconds to initiate the plasma and drive a plasma current up to 2 MA. The poloidal field system consists of 14 individual coils which are arranged symmetrically above and below the horizontal mid plane. Four pairs of coils make up the central solenoid, and three pairs of poloidal ring coils complete the system. The poloidal field coils all use a cable-in-conduit conductor, using either niobium 3-tin (Nb 3 Sn) or niobium titanium (NbTi) superconducting strands depending on the operating conditions for that coil. All of the coils are cooled by flowing supercritical helium, with inlet and outlet connections made on each double pancake. The superconducting magnet system has gone through a conceptual design review, and is in preliminary design started by the LLNL/MIT/PPPL collaboration. A number of changes have been made in the design since the conceptual design review, and are described in this paper. The majority of the design and all fabrication of the superconducting magnet system will be ,accomplished by industry, which will shortly be taking over the preliminary design. The magnet system is expected to be completed in early 2000
Energy Technology Data Exchange (ETDEWEB)
Vahala, George M. [College of William and Mary, Williamsburg, VA (United States)
2013-12-31
Lattice Boltzmann algorithms are a mesoscopic method to solve problems in nonlinear physics which are highly parallelized – unlike the direction solution of the original problem. These methods are applied to both fluid and magnetohydrodynamic turbulence. By introducing entropic constraints one can enforce the positive definiteness of the distribution functions and so be able to simulate fluids at high Reynolds numbers without numerical instabilities. By introducing a vector distribution function for the magnetic field one can enforce the divergence free condition on the magnetic field automatically, without the need of divergence cleaning as needed in most direct numerical solutions of the resistive magnetohydrodynamic equations. The principal reason for the high parallelization of lattice Boltzmann codes is that they consist of a kinetic collisional relaxation step (which is purely local) followed by a simple shift of the relaxed data to neighboring lattice sites. In large eddy simulations, the closure schemes are highly nonlocal – the most famous of these schemes is that due to Smagorinsky. Under a lattice Boltzmann representation the Smagorinsky closure is purely local – being simply a particular moment on the perturbed distribution fucntions. After nonlocal fluid moment models were discovered to represent Landau damping, it was found possible to model these fluid models using an appropriate lattice Boltzmann algorithm. The close to ideal parallelization of the lattice Boltzmann codes permitted us to be Gordon Bell finalists on using the Earth Simulation in Japan. We have also been involved in the radio frequency propagation of waves into a tokamak and into a spherical overdense tokamak plasma. Initially we investigated the use of a quasi-optical grill for the launching of lower hybrid waves into a tokamak. It was found that the conducting walls do not prevent the rods from being properly irradiated, the overloading of the quasi-optical grill is not severe
Trapping of gun-injected plasma by a tokamak
International Nuclear Information System (INIS)
Leonard, A.W.; Dexter, R.N.; Sprott, J.C.
1986-10-01
It is shown that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. Gun injection raises the line-averaged density and peaks the density profile. Trapping of the gun-injected plasma is explainable in terms of a depolarization current mechanism. A model is developed which describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/ ∞ n -1 /sub b/T 3 /sub e/(α 0 /L) 2 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions
Trapping of gun-injected plasma by a tokamak
International Nuclear Information System (INIS)
Leonard, A.W.; Dexter, R.N.; Sprott, J.C.
1987-01-01
It has been seen that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. This trapping of a gun-injected plasma is explained in terms of a depolarization current mechanism. A model is developed that describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/proportionalT/sup 3/2//sub e/L 2 /n/sub b/α 2 0 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions
[Research programs in plasma physics]: Annual report
International Nuclear Information System (INIS)
Weitzner, H.
1988-01-01
This paper contains a brief review of the work done in 1987 at New York University in plasma physics. Topics discussed in this report are: reduction and interpretation of experimental tokamak data, turbulent transport in tokamaks and RFP's, laminar flow transport, wave propagation in different frequency regimes, stability of flows, plasma fueling, magnetic reconnection problems, development of new numerical techniques for Fokker-Planck-like equations, and stability of shock waves. Outside of fusion there has been work in free electron lasers, heating of solar coronal loops and renormalized theory of fluid turbulence
Rippling modes in the edge of a tokamak plasma
International Nuclear Information System (INIS)
Carreras, B.A.; Callen, J.D.; Gaffney, P.W.; Hicks, H.R.
1982-02-01
A promising resistive magnetohydrodynamic candidate for the underlying cause of turbulence in the edge of a tokamak plasma is the rippling instability. In this paper we develop a computational model for these modes in the cylindrical tokamak approximation and explore the linear growth and single-helicity quasi-linear saturation phases of the rippling modes for parameters appropriate to the edge of a tokamak plasma. Large parallel heat conduction does not stabilize these modes; it only reduces their growth rate by a factor scaling as k/sub parallel//sup -4/3/. Nonlinearly, individual rippling modes are found to saturate by quasi-linear flattening of the resistivity profile. The saturated amplitude of the modes scales as m/sup -1/, and the radial extent of these modes grows linearly with time due to radial Vector E x Vector B 0 convection. This evolution is found to be terminated by parallel heat conduction
Rippling modes in the edge of a tokamak plasma
International Nuclear Information System (INIS)
Carreras, B.A.; Gaffney, P.W.; Hicks, H.R.; Callan, J.D.
1982-01-01
A promising resistive magnetohydrodynamic candidate for the underlying cause of turbulence in the edge of a tokamak plasma is the rippling instability. In this paper a computational model for these modes in the cylindrical tokamak approximation was developed and the linear growth and single-helicity quasi-linear saturation phases of the rippling modes for parameters appropriate to the edge of a tokamak plasma were explored. Large parallel heat conduction does not stabilize these modes; it only reduces their growth rate by a factor sacling as K/sup -4/3//sub parallel/. Nonlinearly, individual rippling modes are found to saturate by quasi-linear flattening of the resistivity profile. The saturated amplitude of the modes scales as m -1 , and the radial extent of these modes grows linearly with time due to radial E x B 0 convection. This evolution is found to be terminated by parallel heat conduction
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)
Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta
International Nuclear Information System (INIS)
Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.
2017-01-01
Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t ), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.
Equilibrium of rotating and nonrotating plasmas in tokamaks
International Nuclear Information System (INIS)
Pustovitov, V.D.
2003-01-01
One studied plasma equilibrium in tokamak in case of toroidal rotation. Rotation associated centrifugal force is shown to result in decrease of equilibrium limit as to β. One analyzes unlike opinion and considers its supports. It is shown that in possible case of local improvement of equilibrium conditions associated with special selection of profile of plasma rotation rate, the combined integral effect turns to be negative one. But in case of typical conditions, decrease of equilibrium β caused by plasma rotation is negligible one and one may ignore effect of plasma rotation on its equilibrium for hot plasma [ru
Experimental methods to study tokamak plasma stability
International Nuclear Information System (INIS)
Perez-Navarro, A.
1978-01-01
Experimental devices to measure external instability modes with small pick-up coils to detect poloidal magnetic field fluctuations, and internal modes with soft-X-ray detectors are discussed. The characteristics of these devices are calculated for a small tokamak (R 0 = 30 cm, a = 10 cm, I 0 50 KA). (author)
[High beta tokamak research and plasma theory
International Nuclear Information System (INIS)
1990-01-01
Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report
Anomalous periodic disruptions in tokamak plasma
International Nuclear Information System (INIS)
Montvai, A.; Tegze, M.; Valyi, I.
1982-09-01
Anomalously strong, periodic instabilities were observed in the MT-1 tokamak. Characteristics of these instabilities were partly similar to those of internal disruptions, but there were features making them different from the normal relaxational oscillations. Basic characteristics of the phenomenon were studied with the aid of generally used diagnostics. (author)
Dynamic behavior of plasma-facing materials during plasma instabilities in tokamak reactors
International Nuclear Information System (INIS)
Hassanein, A.; Konkashbaev, I.
1997-01-01
Damage to plasma-facing and nearby components due to plasma instabilities remains a major obstacle to a successful tokamak concept. The high energy deposited on facing materials during plasma instabilities can cause severe erosion, plasma contamination, and structural failure of these components. Erosion damage can take various forms such as surface vaporization, spallation, and liquid ejection of metallic materials. Comprehensive thermodynamic and radiation hydrodynamic codes have been developed, integrated, and used to evaluate the extent of various damage to plasma-facing and nearby components. The eroded and splashed materials will be transported and then redeposited elsewhere on other plasma-facing components. Detailed physics of plasma/solid-liquid/vapor interaction in a strong magnetic field have been developed, optimized, and implemented in a self-consistent model. The plasma energy deposited in the evolving divertor debris is quickly and intensely reradiated, which may cause severe erosion and melting of other nearby components. Factors that influence and reduce vapor-shielding efficiency such as vapor diffusion and turbulence are also discussed and evaluated
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.
Detection of tokamak plasma positrons using annihilation photons
Energy Technology Data Exchange (ETDEWEB)
Guanying, Yu; Liu, Jian; Xie, Jinlin [University of Science and Technology, Hefei, Anhui, 230027 (China); Li, Jiangang, E-mail: j_li@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)
2017-05-15
Highlights: • A design for detection of tokamak plasma positrons is given. • Identify the main obstacle toward experimental confirmation of fusion plasma positrons. • Signal to noise ratio in a plasma disruption is estimated. • Unique potential applications of fusion plasma positrons are discussed. - Abstract: A massive amount of positrons (plasma positrons), produced by the collision between runaway electrons and nuclei during fusion plasma disruption, was first predicted theoretically in 2003. To help confirm this prediction, we report here the design of an experimental system to detect tokamak plasma positrons. Because a substantial amount of positrons (material positrons) are produced when runaway electrons impact plasma-facing materials, we proposed maximizing the ratio of plasma to material positrons by inserting a thin carbon target at the plasma edge as a plasma positron bombing target and producing a plasma disruption scenario triggered by massive gas injection. Meanwhile, the coincidence detection of positron annihilation photons was used to filter out the noise of annihilation photons from locations other than the carbon target and that of bremsstrahlung photons near 511 keV. According to our simulation, the overall signal-to-noise ratio should be more than 10:1.
Energy Technology Data Exchange (ETDEWEB)
Chrystal, C. [University of California-San Diego, La Jolla, California 92186-5608 (United States); Burrell, K. H.; Groebner, R. J.; Kaplan, D. H. [General Atomics, San Diego, California 92186-5608 (United States); Grierson, B. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)
2012-10-15
To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.
Chrystal, C; Burrell, K H; Grierson, B A; Groebner, R J; Kaplan, D H
2012-10-01
To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.
Development in Diagnostics Application to Control Advanced Tokamak Plasma
International Nuclear Information System (INIS)
Koide, Y.
2008-01-01
For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)
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)
Hybrid model for simulation of plasma jet injection in tokamak
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.
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
International Nuclear Information System (INIS)
Federici, G.; Skinner, C.H.; Brooks, J.N.
2001-01-01
The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically in influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the part of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material inter actions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (author)
International Nuclear Information System (INIS)
Federici, G.; Skinner, C.H.; Brooks, J.N.
2001-01-01
The major increase in discharge duration and plasma energy in a next-step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several cm from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the engineering design activities of the international thermonuclear experimental reactor project (ITER) and significant progress has been made in better understanding these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/re-deposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (orig.)
Neural net prediction of tokamak plasma disruptions
International Nuclear Information System (INIS)
Hernandez, J.V.; Lin, Z.; Horton, W.; McCool, S.C.
1994-10-01
The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system
International Nuclear Information System (INIS)
Saito, T.; Hamada, Y.; Yamashita, T.; Ikeda, M.; Nakamura, M.
1980-01-01
The SMM wave laser scattering apparatus has been developed for the measurement of the waves and turbulences in the plasma. This apparatus will help greatly to clarify the physics of RF heating of the tokamak plasma. The present status of main parts of the apparatus, the SMM wave laser and the Schottky barrier diode mixer for the heterodyne receiver, are described. (author)
Plasma physics and engineering
Fridman, Alexander
2011-01-01
Part I: Fundamentals of Plasma Physics and Plasma ChemistryPlasma in Nature, in the Laboratory, and in IndustryOccurrence of Plasma: Natural and Man MadeGas DischargesPlasma Applications, Plasmas in IndustryPlasma Applications for Environmental ControlPlasma Applications in Energy ConversionPlasma Application for Material ProcessingBreakthrough Plasma Applications in Modern TechnologyElementary Processes of Charged Species in PlasmaElementary Charged Particles in Plasma and Their Elastic and Inelastic CollisionsIonization ProcessesMechanisms of Electron Losses: The Electron-Ion RecombinationEl
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
Losses of runaway electrons in MHD-active plasmas of the COMPASS tokamak.
Czech Academy of Sciences Publication Activity Database
Ficker, Ondřej; Mlynář, Jan; Vlainic, Milos; Čeřovský, Jaroslav; Urban, Jakub; Vondráček, Petr; Weinzettl, Vladimír; Macúšová, Eva; Decker, J.; Gospodarczyk, M.; Martin, P.; Nardon, E.; Papp, G.; Plyusnin, V.V.; Reux, C.; Saint-Laurent, F.; Sommariva, C.; Cavalier, Jordan; Havlíček, Josef; Havránek, Aleš; Hronová-Bilyková, Olena; Imríšek, Martin; Markovič, Tomáš; Varju, Jozef; Papřok, Richard; Pánek, Radomír; Hron, Martin
2017-01-01
Roč. 57, č. 7 (2017), č. článku 076002. ISSN 0029-5515 R&D Projects: GA MŠk LG14002; GA MŠk(CZ) LM2015045; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tokamaks * runaway electrons * MHD instabilities * disruptions Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016
Physics of plasma-wall interactions in controlled fusion
International Nuclear Information System (INIS)
Post, D.E.; Behrisch, R.
1984-01-01
In the areas of plasma physics, atomic physics, surface physics, bulk material properties and fusion experiments and theory, the following topics are presented: the plasma sheath; plasma flow in the sheath and presheath of a scrape-off layer; probes for plasma edge diagnostics in magnetic confinement fusion devices; atomic and molecular collisions in the plasma boundary; physical sputtering of solids at ion bombardment; chemical sputtering and radiation enhanced sublimation of carbon; ion backscattering from solid surfaces; implantation, retention and release of hydrogen isotopes; surface erosion by electrical arcs; electron emission from solid surfaces;l properties of materials; plasma transport near material boundaries; plasma models for impurity control experiments; neutral particle transport; particle confinement and control in existing tokamaks; limiters and divertor plates; advanced limiters; divertor tokamak experiments; plasma wall interactions in heated plasmas; plasma-wall interactions in tandem mirror machines; and impurity control systems for reactor experiments
System studies for quasi-steady-state advanced physics tokamak
International Nuclear Information System (INIS)
Reid, R.L.; Peng, Y.K.M.
1983-11-01
Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated
Czech Academy of Sciences Publication Activity Database
Hron, Martin; Ďuran, Ivan; Stöckel, Jan; Hidalgo, C.
2004-01-01
Roč. 54, suppl. C (2004), C22-C27 ISSN 0011-4626. [Symposium on Plasma Physics and Technology /21st/. Praha, 14.06.2004-17.06.2004] R&D Projects: GA ČR GA202/03/0786 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, edge plasma, polarization Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.292, year: 2004
Electron-cyclotron current drive in the tokamak physics experiment
International Nuclear Information System (INIS)
Smith, G.R.; Kritz, A.H.; Radin, S.H.
1992-01-01
Ray-tracking calculations provide estimates of the electron-cyclotron heating (ECH) power required to suppress tearing modes near the q=2 surface in the Tokamak Physics Experiment. Effects of finite beam width and divergence are included, as are the effects of scattering of the ECH power by drift-wave turbulence. A frequency of about 120 GHz allows current drive on the small-R (high-B) portion of q=2, while 80 GHz drives current on the large-R (low-B) portion. The higher frequency has the advantages of less sensitivity to wave and plasma parameters and of no trapped-electron degradation of current-drive efficiency. Less than 1 MW suffices to suppress tearing modes even with high turbulence levels
The cryogenic helium cooling system for the Tokamak physics experiment
International Nuclear Information System (INIS)
Felker, B.; Slack, D.S.; Wendland, C.R.
1995-01-01
The Tokamak Physics Experiment (TPX) will use supercritical helium to cool all the magnets and supply helium to the Vacuum cryopumping subsystem. The heat loads will come from the standard steady state conduction and thermal radiation sources and from the pulsed loads of the nuclear and eddy currents caused by the Central Solenoid Coils and the plasma positioning coils. The operations of the TPX will begin with pulses of up to 1000 seconds in duration every 75 minutes. The helium system utilizes a pulse load leveling scheme to buffer out the effects of the pulse load and maintain a constant cryogenic plant operation. The pulse load leveling scheme utilizes the thermal mass of liquid and gaseous helium stored in a remote dewar to absorb the pulses of the tokamak loads. The mass of the stored helium will buffer out the temperature pulses allowing 5 K helium to be delivered to the magnets throughout the length of the pulse. The temperature of the dewar will remain below 5 K with all the energy of the pulse absorbed. This paper will present the details of the heat load sources, of the pulse load leveling scheme operations, a partial helium schematic, dewar temperature as a function of time, the heat load sources as a function of time and the helium temperature as a function of length along the various components that will be cooled
An introduction to boundary plasma physics
International Nuclear Information System (INIS)
Shimizu, Katsuhiro; Takizuka, Tomonori
2004-01-01
History of tokamak experiments is briefly reviewed with a special focus on divertors. Two-point divertor model, which calculates plasma parameters up-stream and at the divertor plate for a given condition of particle flux and heat flux, is explained. The model is applied to ITER to discuss the heat flux onto the target plate. The important issues of divertor physics related to recycling, remote radiative cooling, detached plasma and MARFE are also introduced. (author)
Plasma shaping effects on tokamak scrape-off layer turbulence
Riva, Fabio; Lanti, Emmanuel; Jolliet, Sébastien; Ricci, Paolo
2017-03-01
The impact of plasma shaping on tokamak scrape-off layer (SOL) turbulence is investigated. The drift-reduced Braginskii equations are written for arbitrary magnetic geometries, and an analytical equilibrium model is used to introduce the dependence of turbulence equations on tokamak inverse aspect ratio (ε ), Shafranov’s shift (Δ), elongation (κ), and triangularity (δ). A linear study of plasma shaping effects on the growth rate of resistive ballooning modes (RBMs) and resistive drift waves (RDWs) reveals that RBMs are strongly stabilized by elongation and negative triangularity, while RDWs are only slightly stabilized in non-circular magnetic geometries. Assuming that the linear instabilities saturate due to nonlinear local flattening of the plasma gradient, the equilibrium gradient pressure length {L}p=-{p}e/{{\
Pseudo-MHD ballooning modes in tokamak plasmas
International Nuclear Information System (INIS)
Callen, J.D.; Hegna, C.C.
1996-08-01
The MHD description of a plasma is extended to allow electrons to have both fluid-like and adiabatic-regime responses within an instability eigenmode. In the resultant open-quotes pseudo-MHDclose quotes model, magnetic field line bending is reduced in the adiabatic electron regime. This makes possible a new class of ballooning-type, long parallel extent, MHD-like instabilities in tokamak plasmas for α > s 2 (2 7/3 /9) (r p /R 0 ) or-d√Β/dr > (2 1/6 /3)(s/ R 0q ), which is well below the ideal-MHD stability boundary. The marginally stable pressure profile is similar in both magnitude and shape to that observed in ohmically heated tokamak plasmas
Heating of plasmas in tokamaks by current-driven turbulence
International Nuclear Information System (INIS)
Kluiver, H. de.
1985-10-01
Investigations of current-driven turbulence have shown the potential to heat plasmas to elevated temperatures in relatively small cross-section devices. The fundamental processes are rather well understood theoretically. Even as it is shown to be possible to relax the technical requirements on the necessary electric field and the pulse length to acceptable values, the effect of energy generation near the plasma edge, the energy transport, the impurity influx and the variation of the current profile are still unknown for present-day large-radius tokamaks. Heating of plasmas by quasi-stationary weakly turbulent states caused by moderate increases of the resistivity due to higher loop voltages could be envisaged. Power supplies able to furnish power levels 5-10 times higher than the usual values could be used for a demonstration of those regimes. At several institutes and university laboratories the study of turbulent heating in larger tokamaks and stellarators is pursued
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
Stacey, Weston M
2012-01-01
This revised and enlarged second edition of the popular textbook and reference contains comprehensive treatments of both the established foundations of magnetic fusion plasma physics and of the newly developing areas of active research. It concludes with a look ahead to fusion power reactors of the future. The well-established topics of fusion plasma physics -- basic plasma phenomena, Coulomb scattering, drifts of charged particles in magnetic and electric fields, plasma confinement by magnetic fields, kinetic and fluid collective plasma theories, plasma equilibria and flux surface geometry, plasma waves and instabilities, classical and neoclassical transport, plasma-materials interactions, radiation, etc. -- are fully developed from first principles through to the computational models employed in modern plasma physics. The new and emerging topics of fusion plasma physics research -- fluctuation-driven plasma transport and gyrokinetic/gyrofluid computational methodology, the physics of the divertor, neutral ...
Automation of Aditya tokamak plasma position control DC power supply
Energy Technology Data Exchange (ETDEWEB)
Arambhadiya, Bharat, E-mail: bharat@ipr.res.in; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.
2016-11-15
Highlights: • Plasma position control is very essential for obtaining repeatable high temperature, high-density discharges of longer durations in tokomak. • The present capacitor bank has limitations of maximum current capacity and position control beyond 200 ms. • The installation of a separate set of coils and a DC power supply can control the plasma position beyond 200 ms. • A high power thyristor (T588N1200) triggers for DC current pulse of 300 A fires precisely at required positions to modify plasma position. • The commissioning is done for the automated in-house, quick and reliable solution. - Abstract: Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200 ms for position control. This motivated to have a DC power supply of 500 A to provide current pulse beyond 200 ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.
Energy Technology Data Exchange (ETDEWEB)
Martynenko, Yu. V., E-mail: Martynenko-YV@nrcki.ru [National Research Nuclear University “MEPhI” (Russian Federation)
2017-03-15
It is shown that the shielding plasma layer and metal droplet erosion in tokamaks are closely interrelated, because shielding plasma forms from the evaporated metal droplets, while droplet erosion is caused by the shielding plasma flow over the melted metal surface. Analysis of experimental data and theoretical models of these processes is presented.
'Snowflake' H Mode in a Tokamak Plasma
International Nuclear Information System (INIS)
Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.
2010-01-01
An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a 'snowflake' (second-order null) divertor configuration for the first time in the TCV tokamak. This regime exhibits 2 to 3 times lower ELM frequency and 20%-30% increased normalized ELM energy (ΔW ELM /W p ) compared to an identically shaped, conventional single-null diverted H mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.
Plasma radiation in tokamak disruption simulation experiments
International Nuclear Information System (INIS)
Arkhipov, N.; Bakhtin, V.; Safronov, V.; Toporkov, D.; Vasenin, S.; Zhitlukhin, A.; Wuerz, H.
1995-01-01
Plasma impact results in sudden evaporation of divertor plate material and produces a plasma cloud which acts as a protective shield. The incoming energy flux is absorbed in the plasma shield and is converted mainly into radiation. Thus the radiative characteristics of the target plasma determine the dissipation of the incoming energy and the heat load at the target. Radiation of target plasma is studied at the two plasma gun facility 2MK-200 at Troitsk. Space- and time-resolved spectroscopy and time-integrated space-resolved calorimetry are employed as diagnostics. Graphite and tungsten samples are exposed to deuterium plasma streams. It is found that the radiative characteristics depend strongly on the target material. Tungsten plasma arises within 1 micros close to the surface and shows continuum radiation only. Expansion of tungsten plasma is restricted. For a graphite target the plasma shield is a mixture of carbon and deuterium. It expands along the magnetic field lines with a velocity of v = (3--4) 10 6 cm/s. The plasma shield is a two zone plasma with a hot low dense corona and a cold dense layer close to the target. The plasma corona emits intense soft x-ray (SXR) line radiation in the frequency range from 300--380 eV mainly from CV ions. It acts as effective dissipation system and converts volumetrically the incoming energy flux into SXR radiation
International Nuclear Information System (INIS)
Rubenchik, A.; Witkowski, S.
1991-01-01
This book provides a comprehensive review of laser fusion plasma physics and contains the most up-to-date information on high density plasma physics and radiation transport, useful for astrophysicists and high density physicists
Sawtooth driven particle transport in tokamak plasmas
International Nuclear Information System (INIS)
Nicolas, T.
2013-01-01
The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr
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)
Initial plasma production by induction electric field on QUEST tokamak
International Nuclear Information System (INIS)
Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke
2007-01-01
Induction electric field by center solenoid coil plays a roll to produce initial plasma. According to Townsend avalanche theory, minimum electric field for plasma breakdown depends on neutral gas pressure and connection length. On QUEST spherical tokamak, a connection length is evaluated as 966m on null point neighborhood with coil current ratio I PF26 /I CS =0.1, and induction electric field considering eddy current of vacuum vessel is evaluated as about 0.1 V/m on null point neighborhood. With Townsend avalanche theory, these values manage to produce initial plasma on QUEST. (author)
A midsize tokamak as a fast track to burning plasmas
Directory of Open Access Journals (Sweden)
E. Mazzucato
2011-03-01
Full Text Available This paper describes the conceptual design of a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥ 10 with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER. This can be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a different magnetic divertor from those currently employed in present experiments is discussed.
Wave trajectory and electron cyclotron heating in tokamak plasmas
International Nuclear Information System (INIS)
Tanaka, S.; Maekawa, T.; Terumichi, Y.; Hamada, Y.
1980-01-01
Wave trajectories in high density tokamak plasmas are studied numerically. Results show that the ordinary wave injected at an appropriate incident angle can propagate into the dense plasmas and is mode-converted to the extraordinary wave at the plasma cutoff, is further converted to the electron Bernstein wave during passing a loop or a folded curve near the upper hybrid resonance layer, and is cyclotron damped away, resulting in local electron heating before arriving at the cyclotron resonance layer. Similar trajectory and damping are obtained when a microwave in a form of extraordinary wave is injected quasi-perpendicularly in the direction of decreasing toroidal field
Plasma physics an introduction
Fitzpatrick, Richard
2014-01-01
Plasma Physics: An Introduction is based on a series of university course lectures by a leading name in the field, and thoroughly covers the physics of the fourth state of matter. This book looks at non-relativistic, fully ionized, nondegenerate, quasi-neutral, and weakly coupled plasma. Intended for the student market, the text provides a concise and cohesive introduction to plasma physics theory, and offers a solid foundation for students wishing to take higher level courses in plasma physics.
Scrape-off layer plasma modeling for the DIII-D tokamak
International Nuclear Information System (INIS)
Porter, G.D.; Rognlien, T.D.; Allen, S.L.
1994-09-01
The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing
Control strategy for plasma equilibrium in a tokamak
International Nuclear Information System (INIS)
Miskell, R.V.
1975-01-01
The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)
MHD Effects of a Ferritic Wall on Tokamak Plasmas
Hughes, Paul E.
It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency
International Nuclear Information System (INIS)
Berezhnyj, V.L.; Kononenko, V.I.; Epishin, V.A.; Topkov, A.N.
1992-01-01
The review of interferometric methods of plasma investigation in the wave submillimeter range is given. The diagnostic schemes in stellarators and tokamaks designed for experienced thermonuclear reactors and also the perspective ones, which are still out of practice, are shown. The methods of these diagnostics, their physical principles, the main possibilities and restrictions at changes of electron density, magnetic fields (currents) and their spatial distributions are described. 105 refs.; 9 figs.; 2 tables. (author)
Fukuda, Takeshi
The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.
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
Wills Plasma Physics Department annual report, 1989
International Nuclear Information System (INIS)
1991-01-01
An overview of the collaborative researches carried out during the 1989 at the Wills Plasma Physics Department is given. The main activities included the study of hydromagnetic surface waves and RF heating using the Tortus tokamak; the development of diagnostic techniques, particularly those based on submillimetre lasers and tunable gyrotrons; gas discharge studies and investigations of apparent cold nuclear fusion in deuterated palladium. The small research tokamak Tortus was upgraded during the year, thus enabling the machine to be routinely and reliably operated at toroidal currents around 40 kA. A list of papers published or presented at various conferences during the year is included in the Appendix
Locked magnetic island chains in toroidally flow damped tokamak plasmas
International Nuclear Information System (INIS)
Fitzpatrick, R; Waelbroeck, F L
2010-01-01
The physics of a locked magnetic island chain maintained in the pedestal of an H-mode tokamak plasma by a static, externally generated, multi-harmonic, helical magnetic perturbation is investigated. The non-resonant harmonics of the external perturbation are assumed to give rise to significant toroidal flow damping in the pedestal, in addition to the naturally occurring poloidal flow damping. Furthermore, the flow damping is assumed to be sufficiently strong to relax the pedestal ion toroidal and poloidal fluid velocities to fixed values determined by neoclassical theory. The resulting neoclassical ion flow causes a helical phase-shift to develop between the locked island chain and the resonant harmonic of the external perturbation. Furthermore, when this phase-shift exceeds a critical value, the chain unlocks from the resonant harmonic and starts to rotate, after which it decays away and is replaced by a helical current sheet. The neoclassical flow also generates an ion polarization current in the vicinity of the island chain which either increases or decreases the chain's radial width, depending on the direction of the flow. If the polarization effect is stabilizing, and exceeds a critical amplitude, then the helical island equilibrium becomes unstable, and the chain again decays away. The critical amplitude of the resonant harmonic of the external perturbation at which the island chain either unlocks or becomes unstable is calculated as a function of the pedestal ion pressure, the neoclassical poloidal and toroidal ion velocities and the poloidal and toroidal flow damping rates.
Understanding L-H transition in tokamak fusion plasmas
Xu, Guosheng; Wu, Xingquan
2017-03-01
This paper reviews the current state of understanding of the L-H transition phenomenon in tokamak plasmas with a focus on two central issues: (a) the mechanism for turbulence quick suppression at the L-H transition; (b) the mechanism for subsequent generation of sheared flow. We briefly review recent advances in the understanding of the fast suppression of edge turbulence across the L-H transition. We uncover a comprehensive physical picture of the L-H transition by piecing together a number of recent experimental observations and insights obtained from 1D and 2D simulation models. Different roles played by diamagnetic mean flow, neoclassical-driven mean flow, turbulence-driven mean flow, and turbulence-driven zonal flows are discussed and clarified. It is found that the L-H transition occurs spontaneously mediated by a shift in the radial wavenumber spectrum of edge turbulence, which provides a critical evidence for the theory of turbulence quench by the flow shear. Remaining questions and some key directions for future investigations are proposed. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contracts No. 2015GB101000, No. 2013GB106000, and No. 2013GB107000 and National Natural Science Foundation of China under Contracts No. 11575235 and No. 11422546.
Two-fluid and parallel compressibility effects in tokamak plasmas
International Nuclear Information System (INIS)
Sugiyama, L.E.; Park, W.
1998-01-01
The MHD, or single fluid, model for a plasma has long been known to provide a surprisingly good description of much of the observed nonlinear dynamics of confined plasmas, considering its simple nature compared to the complexity of the real system. On the other hand, some of the supposed agreement arises from the lack of the detailed measurements that are needed to distinguish MHD from more sophisticated models that incorporate slower time scale processes. At present, a number of factors combine to make models beyond MHD of practical interest. Computational considerations still favor fluid rather than particle models for description of the full plasma, and suggest an approach that starts from a set of fluid-like equations that extends MHD to slower time scales and more accurate parallel dynamics. This paper summarizes a set of two-fluid equations for toroidal (tokamak) geometry that has been developed and tested as the MH3D-T code [1] and some results from the model. The electrons and ions are described as separate fluids. The code and its original MHD version, MH3D [2], are the first numerical, initial value models in toroidal geometry that include the full 3D (fluid) compressibility and electromagnetic effects. Previous nonlinear MHD codes for toroidal geometry have, in practice, neglected the plasma density evolution, on the grounds that MHD plasmas are only weakly compressible and that the background density variation is weaker than the temperature variation. Analytically, the common use of toroidal plasma models based on aspect ratio expansion, such as reduced MHD, has reinforced this impression, since this ordering reduces plasma compressibility effects. For two-fluid plasmas, the density evolution cannot be neglected in principle, since it provides the basic driving energy for the diamagnetic drifts of the electrons and ions perpendicular to the magnetic field. It also strongly influences the parallel dynamics, in combination with the parallel thermal
MTX [Microwave Tokamak Experiment] plasma diagnostic system
International Nuclear Information System (INIS)
Rice, B.W.; Hooper, E.B.; Brooksby, C.A.
1987-01-01
In this paper, a general overview of the MTX plasma diagnostics system is given. This includes a description of the MTX machine configuration and the overall facility layout. The data acquisition system and techniques for diagnostic signal transmission are also discussed. In addition, the diagnostic instruments planned for both an initial ohmic-heating set and a second FEL-heating set are described. The expected range of plasma parameters along with the planned plasma measurements will be reviewed. 7 refs., 5 figs
Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment
Lucia, Matthew James
The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (dLi ˜ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2 O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H 2 within minutes. For impurity sequestration, LTX plasma performance
Numerical studies of transport processes in Tokamak plasma
International Nuclear Information System (INIS)
Spineanu, F.; Vlad, M.
1984-09-01
The paper contains the summary of a set of studies of the transport processes in tokamak plasma, performed with a one-dimensional computer code. The various transport models (which are implemented by the expressions of the transport coefficients) are presented in connection with the regimes of the dynamical development of the discharge. Results of studies concerning the skin effect and the large scale MHD instabilities are also included
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
International Nuclear Information System (INIS)
Castracane, J.
2001-01-01
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies
MAIA, Eigenvalues for MHD Equation of Tokamak Plasma Stability Problems
International Nuclear Information System (INIS)
Tanaka, Y.; Azumi, M.; Kurita, G.; Tsunematsu, T.; Takeda, T.
1986-01-01
1 - Description of program or function: This program solves an eigenvalue problem zBx=Ax where A and B are real block tri-diagonal matrices. This eigenvalue problem is derived from a reduced set of linear resistive MHD equations which is often employed to study tokamak plasma stability problem. 2 - Method of solution: Both the determinant and inverse iteration methods are employed. 3 - Restrictions on the complexity of the problem: The eigenvalue z must be real
Momentum Injection in Tokamak Plasmas and Transitions to Reduced Transport
International Nuclear Information System (INIS)
Parra, F. I.; Highcock, E. G.; Schekochihin, A. A.; Barnes, M.; Cowley, S. C.
2011-01-01
The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.
Plasma current profile during current reversal in a tokamak
International Nuclear Information System (INIS)
Huang Jianguo; Yang Xuanzong; Zheng Shaobai; Feng Chunhua; Zhang Houxian; Wang Long
1999-01-01
Alternating current operation with one full cycle and a current level of 2.5 kA have been achieved in the CT-6B tokamak. The poloidal magnetic field in the plasma is measured with two internal magnetic probes in repeated discharges. The current distribution is reconstructed with an inversion algorithm. The inverse current first appears on the weak field side. The existence of magnetic surfaces and rotational transform provide particle confinement in the current reversal phase
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Castracane, J.
2001-01-04
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.
Burn stability of tokamak fusion plasmas with synergetic current drive
International Nuclear Information System (INIS)
Anderson, D.; Lisak, M.; Kolesnichenko, Ya.
1991-01-01
The stability of thermonuclear burn in Tokamak-reactors with non-inductive current generated with the simultaneous application of various methods is investigated. Particular emphasis is given to the ITER synergetic current drive scenario involving LH waves, neoclassical effects and NB injection. For ITER-like confinement laws, it is shown that this scenario may be unstable on the plasma skin time scale. Figs
Princeton University Plasma Physics Laboratory, Princeton, New Jersey
International Nuclear Information System (INIS)
1991-01-01
This report discusses the following topics: Principal parameters of experimental devices; Tokamak Fusion Test Reactor; Burning Plasma Experiment; Princeton Beta Experiment-Modification; Current Drive Experiment-Upgrade; International Thermonuclear Experimental Reactor; International Collaboration; X-Ray Laser Studies; Hyperthermal Atomic Beam Source; Pure Electron Plasma Experiments; Plasma Processing: Deposition and Etching of Thin Films; Theoretical Studies; Tokamak Modeling; Engineering Department; Environment, Safety, and Health and Quality Assurance; Technology Transfer; Office of Human Resources and Administration; PPPL Patent Invention Disclosures; Office of Resource Management; Graduate Education: Plasma Physics; Graduate Education: Program in Plasma Science and Technology; and Science Education Program
Tokamak plasma current disruption infrared control system
International Nuclear Information System (INIS)
Kugel, H.W.; Ulrickson, M.
1987-01-01
This patent describes a device for magnetically confining a plasma driven by a plasma current and contained within a toroidal vacuum chamber, the device having an inner toroidal limiter on an inside wall of the vacuum chamber and an arrangement for the rapid prediction and control in real time of a major plasma disruption. The arrangement is described which includes: scanning means sensitive to infrared radiation emanating from within the vacuum chamber, the infrared radiation indicating the temperature along a vertical profile of the inner toroidal limiter. The scanning means is arranged to observe the infrared radiation and to produce in response thereto an electrical scanning output signal representative of a time scan of temperature along the vertical profile; detection means for analyzing the scanning output signal to detect a first peaked temperature excursion occurring along the profile of the inner toroidal limiter, and to produce a detection output signal in repsonse thereto, the detection output signal indicating a real time prediction of a subsequent major plasma disruption; and plasma current reduction means for reducing the plasma current driving the plasma, in response to the detection output signal and in anticipation of a subsequent major plasma disruption
Plasma diagnostics for tokamaks and stellarators
Energy Technology Data Exchange (ETDEWEB)
Stott, P E; Sanchez, J
1994-07-01
A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: Magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Scattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma. (Author) 451 refs.
Plasma diagnostics for tokamaks and stellarators
International Nuclear Information System (INIS)
Stott, P.E.; Sanchez, J.
1994-01-01
A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Sattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma
DIII-D Integrated plasma control solutions for ITER and next-generation tokamaks
International Nuclear Information System (INIS)
Humphreys, D.A.; Ferron, J.R.; Hyatt, A.W.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; In, Y.
2008-01-01
Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed 'integrated plasma control,' enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple ('control-level') models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks
Impact of magnetic perturbation fields on tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Fietz, Sina; Maraschek, Marc; Suttrop, Wolfgang; Zohm, Hartmut [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Classen, Ivo [FOM-Institute DIFFER, Nieuwegein (Netherlands); Collaboration: the ASDEX Upgrade Team
2015-05-01
Non-axisymmetric external magnetic perturbation (MP) fields arise in every tokamak e.g. due to not perfectly positioned external coils. Additionally many tokamaks, like ASDEX Upgrade (AUG), are equipped with a set of external coils, which produce a 3D MP field in addition to the equilibrium field. This field is used to either compensate for the intrinsic MP field or to influence MHD instabilities such as Edge Localised Modes (ELMs) or Neoclassical Tearing Modes (NTMs). But these MP fields can also give rise to a more global plasma response. The resonant components can penetrate the plasma and influence the stability of existing NTMs or even lead to their formation via magnetic reconnection. In addition they exert a local torque on the plasma. These effects are less pronounced at high plasma rotation where the resonant field components are screened. The non-resonant components do not influence NTMs directly but slow down the plasma rotation globally via the neoclassical toroidal viscous torque. The island formation caused by the MP field as well as the interaction of pre-existing islands with the MP field at AUG is presented. It is shown that these effects can be modelled using a simple forced reconnection theory. Also the effect of resonant and non-resonant MPs on the plasma rotation at AUG is discussed.
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
International Nuclear Information System (INIS)
Evans, T.E.
1984-09-01
The first direct observation of the internal structure of driven global Alfven eigenmodes in a tokamak plasma is presented. A carbon dioxide laser scattering/interferometer has been designed, built, and installed on the PRETEXT tokamak. By using this diagnostic system in the interferometer configuration, we have for the first time, thoroughly investigated the resonance conditions required for, and the spatial wave field structure of, driven plasma eigenmodes at frequencies below the ion cyclotron frequency in a confined, high temperature, tokamak plasma
Liquid gallium jet-plasma interaction studies in ISTTOK tokamak
International Nuclear Information System (INIS)
Gomes, R.B.; Fernandes, H.; Silva, C.; Sarakovskis, A.; Pereira, T.; Figueiredo, J.; Carvalho, B.; Soares, A.; Duarte, P.; Varandas, C.; Lielausis, O.; Klyukin, A.; Platacis, E.; Tale, I.; Alekseyv, A.
2009-01-01
Liquid metals have been pointed out as a suitable solution to solve problems related to the use of solid walls submitted to high power loads allowing, simultaneously, an efficient heat exhaustion process from fusion devices. The most promising candidate materials are lithium and gallium. However, lithium has a short liquid state temperature range when compared with gallium. To explore further this property, ISTTOK tokamak is being used to test the interaction of a free flying liquid gallium jet with the plasma. ISTTOK has been successfully operated with this jet without noticeable discharge degradation and no severe effect on the main plasma parameters or a significant plasma contamination by liquid metal. Additionally the response of an infrared sensor, intended to measure the jet surface temperature increase during its interaction with the plasma, has been studied. The jet power extraction capability is extrapolated from the heat flux profiles measured in ISTTOK plasmas.
Impurity screening of scrape-off plasma in a tokamak
International Nuclear Information System (INIS)
Kishimoto, Hiroshi; Tani, Keiji; Nakamura, Hiroo
1981-11-01
Impurity screening effect of a scrape-off layer has been studied in a tokamak, based on a simple model of wall-released impurity behavior. Wall-sputtered impurities are stopped effectively by the scrape-off plasma for a medium-Z or high-Z wall system while major part of impurities enters the main plasma in a low-Z wall system. The screening becomes inefficient with increase of scrape-off plasma temperature. Successive multiplication of recycling impurities in the scrape-off layer is large for a high-Z wall and is enhanced by a rise of scrape-off plasma temperature. The stability of plasma-wall interaction is determined by a multiplication factor of recycling impurities. (author)
Development of a tokamak plasma optimized for stability and confinement
International Nuclear Information System (INIS)
Politzer, P.A.
1995-02-01
Design of an economically attractive tokamak fusion reactor depends on producing steady-state plasma operation with simultaneous high energy density (β) and high energy confinement (τ E ); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H≡ τ E /τ ITER-89P = 4) and high normalized β (β N ≡ β/(I/aB) = 6%-m-T/MA) have been obtained. For the present, these conditions have been produced separately and in transient discharges. The DIII-D advanced tokamak development program is directed toward developing an understanding of the characteristics which lead to high stability and confinement, and to use that understanding to demonstrate stationary, high performance operation through active control of the plasma shape and profiles. The authors have identified some of the features of the operating modes in DIII-D that contribute to better performance. These are control of the plasma shape, control of both bulk plasma rotation and shear in the rotation and Er profiles, and particularly control of the toroidal current profiles. In order to guide their future experiments, they are developing optimized scenarios based on their anticipated plasma control capabilities, particularly using fast wave current drive (on-axis) and electron cyclotron current drive (off-axis). The most highly developed model is the second-stable core VH-mode, which has a reversed magnetic shear safety factor profile [q(O) = 3.9, q min = 2.6, and q 95 = 6]. This model plasma uses profiles which the authors expect to be realizable. At β N ≥ 6, it is stable to n=l kink modes and ideal ballooning modes, and is expected to reach H ≥ 3 with VH-mode-like confinement
Department of Plasma Physics and Technology - Overview
International Nuclear Information System (INIS)
Rabinski, M.
2008-01-01
Full text: The activities of the Department in 2007 continued previous studies in the following fields of plasma physics, controlled nuclear fusion and plasma technology of surface engineering: · Studies of physical phenomena in pulsed discharges in the Plasma-Focus (PF) and RPI-IBIS facilities; · Development of selected methods for high-temperature plasma diagnostics; · Research on plasma technologies; · Selected problems of plasma theory and computational modelling. As for the experimental studies particular attention was paid to the analysis of the correlation of X-ray pulses with pulsed electron beams and other corpuscular emissions from different Plasma-Focus (PF) facilities. A collisional-radiative model, taking into account the Stark effect and strong electric fields in the so called '' hot- spot '' regions of a pinch, was applied in those analyses. The main aim of these studies was to identify the physical phenomena responsible for the emission during the PF-type discharges. The emitted protons were also measured with nuclear track detectors. The measurements made it possible to obtain images of the regions, where the D-D fusion reactions occurred, as well as to determine the angular distribution of the emitted protons. Pulsed plasma streams were also investigated by means of time-resolved optical spectroscopy and corpuscular diagnostics. In a frame of the EURATOM program, efforts were devoted to the development of diagnostic methods for tokamak-type facilities. Such studies include the design and construction of the 4-channel Cherenkov-type detection system for the TORE-SUPRA tokamak at CEA-Cadarache. In the meantime in order to collect some experience a new measuring head was especially prepared for experiments within small facilities. Other fusion- oriented efforts are connected with the application of the solid-state nuclear track detectors for investigation of protons from tokamak plasma and high-energy beams emitted from laser produced plasmas
The role of high speed photography in plasma instability research on the AEC tokamak
International Nuclear Information System (INIS)
Fletcher, J.D.; Coster, D.P.; De Villiers, J.A.M.; Kotze, P.B.; Nothnagel, G.; O'Mahony, J.R.; Roberts, D.E.; Sherwell, D.
1986-01-01
High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions in fusion research devices like tokamaks. Such a system has been installed on the AEC tokamak. This paper reports some preliminary results obtained during typical plasma discharges
Pellet-plasma interactions in tokamaks
DEFF Research Database (Denmark)
Chang, C.T.
1991-01-01
confinement time, offset by the accumulation of impurities at the plasma core is brought into focus. A possible remedy is suggested to diminish the effect of the impurities. Plausible arguments are presented to explain the apparent controversial observations on the propagation of a fast cooling front ahead......The ablation of a refuelling pellet of solid hydrogen isotopes is governed by the plasma state, especially the density and energy distribution of the electrons. On the other hand, the cryogenic pellet gives rise to perturbations of the plasma temperature and density. Based on extensive experimental...... data, the interaction between the pellet and the plasma is reviewed. Among the subjects discussed are the MHD activity, evolution of temperature and density profiles, and the behaviour of impurities following the injection of a pellet (or pellets). The beneficial effect of density peaking on the energy...
Electron cyclotron emission imaging in tokamak plasmas
Munsat, T.; Domier, C.W.; Kong, X. Y.; Liang, T. R.; N C Luhmann Jr.,; Tobias, B. J.; Lee, W.; Park, H. K.; Yun, G.; Classen, I.G.J.; Donne, A. J. H.
2010-01-01
We discuss the recent history and latest developments of the electron cyclotron emission imaging diagnostic technique, wherein electron temperature is measured in magnetically confined plasmas with two-dimensional spatial resolution. The key enabling technologies for this technique are the
Electron temperature gradient driven instability in the tokamak boundary plasma
International Nuclear Information System (INIS)
Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.
1992-01-01
A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t -1/2 e γmt
Plasma facing components design of KT-2 tokamak
International Nuclear Information System (INIS)
In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin
1997-04-01
The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs
Low temperature plasma near a tokamak reactor limiter
International Nuclear Information System (INIS)
Braams, B.J.; Singer, C.E.
1985-01-01
Analytic and two-dimensional computational solutions for the plasma parameters near a toroidally symmetric limiter are illustrated for the projected parameters of a Tokamak Fusion Core Experiment (TFCX). The temperature near the limiter plate is below 20 eV, except when the density 10 cm inside the limiter contact is 8 x 10 13 cm -3 or less and the thermal diffusivity in the edge region is 2 x 10 4 cm 2 /s or less. Extrapolation of recent experimental data suggests that neither of these conditions is likely to be met near ignition in TFCX, so a low plasma temperature near the limiter should be considered a likely possibility
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)
Influence of the helical resonant fields on the plasma potential in the TBR-1 Tokamak
International Nuclear Information System (INIS)
Ribeiro, C.; Silva, R.P. da; Caldas, I.L.; Fagundes, A.N.; Sanada, E.K.
1990-01-01
This work describes an experimental work that are in progress in TBR-1 tokamak about the influence of resonant helical fields on the plasma potential. TBR-1 is a small tokamak in operation in the Physics Institute of University of Sao Paulo and used for basic research, diagnostic development and personal formation. Its main parameters are: R(Major Radius) = 0.30 m; a v (Vessel Radius) = 0.11 m; a(Plasma Radius) = 0.08 m; R/a(Aspect Ratio) = 3.75; B φ (Toroidal Field) = 5 kG; n e0 (Central Electron Density) ≅ 7 x 10 18 m -3 ; T e0 (central electron temperature) ≅ 200 eV. (Author)
International Nuclear Information System (INIS)
Sitenko, A.
1993-01-01
Problems of modern physics and the situation with physical research in Ukraine are considered. Programme of the conference includes scientific and general problems. Its proceeding are published in 6 volumes. The papers presented in this volume refer to plasma physics
International Nuclear Information System (INIS)
Sodha, M.S.; Tewari, D.P.; Subbarao, D.
1983-01-01
The book consists of review articles on some selected contemporary aspects of plasma physics. The selected topics present a panoramic view of contemporary plasma physics and applications to fusion, space and MHD power generation. Basic non-linear plasma theory is also covered. The book is supposed to be useful for M.S./M.Sc. students specialising in plasma physics and for those beginning research work in plasma physics. It will also serve as a valuable reference book for more advanced research workers. (M.G.B.)
Status of plasma physics research activities in Egypt
International Nuclear Information System (INIS)
Masoud, M.M.
1997-01-01
The status of plasma physics research activities in Egypt is reviewed. There are nine institutes with plasma research activities. The largest is the Atomic energy Authority (AEA), which has activities in fundamental plasma studies, fusion technology, plasma and laser applications, and plasma simulation. The experiments include Theta Pinches, a Z Pinch, a coaxial discharge, a glow discharge, a CO 2 laser, and the EGYPTOR tokamak. (author)
Contributions to the 20. EPS conference on controlled fusion and plasma physics
Energy Technology Data Exchange (ETDEWEB)
NONE
1993-07-15
The Conference covers research on different aspects of plasma physics and fusion technology, like technical aspects of Tokamak devices; plasma instabilities and impurities, development and testing of materials for fusion reactors etc.
Vacuum physics analysis of HT-7 superconducting tokamak pump limiter
International Nuclear Information System (INIS)
Hu Jiansheng; Li Chengfu; He Yexi
1998-10-01
The pump limiter is analysed with HT-7 superconducting tokamak parameter and the pump limiter construction. The particle exhaust of the pump limiter can be to achieve about 7.7%. So the pump limiter can be applied in the HT-7 device and will make good affection in plasma discharge
Comments on experimental results of energy confinement of tokamak plasmas
International Nuclear Information System (INIS)
Chu, T.K.
1989-04-01
The results of energy-confinement experiments on steady-state tokamak plasmas are examined. For plasmas with auxiliary heating, an analysis based on the heat diffusion equation is used to define heat confinement time (the incremental energy confinement time). For ohmically sustained plasmas, experiments show that the onset of the saturation regime of energy confinement, marfeing, detachment, and disruption are marked by distinct values of the parameter /bar n//sub e///bar j/. The confinement results of the two types of experiments can be described by a single surface in 3-dimensional space spanned by the plasma energy, the heating power, and the plasma density: the incremental energy confinement time /tau//sub inc/ = ΔW/ΔP is the correct concept for describing results of heat confinement in a heating experiment; the commonly used energy confinement time defined by /tau//sub E/ = W/P is not. A further examination shows that the change of edge parameters, as characterized by the change of the effective collision frequency ν/sub e/*, governs the change of confinement properties. The totality of the results of tokamak experiments on energy confinement appears to support a hypothesis that energy transport is determined by the preservation of the pressure gradient scale length. 70 refs., 6 figs., 1 tab
Twentyseventh European physical society conference on controlled fusion and plasma physics
International Nuclear Information System (INIS)
Igitkhanov, Y.
2000-01-01
The twentyseventh European physical society conference on controlled fusion and plasma physics was held in Budapest, 12-16 June 2000. About 10 invited papers were presented, covering a wide range of problems in plasma physics, including confinement and transport issues in fusion devices, astrophysics and industrial application of plasmas. More than 100 papers were presented on plasma theory and experiments from tokamaks and stellarators. Some of the ITER-relevant issues covered are described in this newsletter
Tokamak plasma variations under rapid compression
International Nuclear Information System (INIS)
Holmes, J.A.; Peng, Y.K.M.; Lynch, S.J.
1980-04-01
Changes in plasmas undergoing large, rapid compressions are examined numerically over the following range of aspect ratios A:3 greater than or equal to A greater than or equal to 1.5 for major radius compressions of circular, elliptical, and D-shaped cross sections; and 3 less than or equal to A less than or equal to 6 for minor radius compressions of circular and D-shaped cross sections. The numerical approach combines the computation of fixed boundary MHD equilibria with single-fluid, flux-surface-averaged energy balance, particle balance, and magnetic flux diffusion equations. It is found that the dependences of plasma current I/sub p/ and poloidal beta anti β/sub p/ on the compression ratio C differ significantly in major radius compressions from those proposed by Furth and Yoshikawa. The present interpretation is that compression to small A dramatically increases the plasma current, which lowers anti β/sub p/ and makes the plasma more paramagnetic. Despite large values of toroidal beta anti β/sub T/ (greater than or equal to 30% with q/sub axis/ approx. = 1, q/sub edge/ approx. = 3), this tends to concentrate more toroidal flux near the magnetic axis, which means that a reduced minor radius is required to preserve the continuity of the toroidal flux function F at the plasma edge. Minor radius compressions to large aspect ratio agree well with the Furth-Yoshikawa scaling laws
International Nuclear Information System (INIS)
Boozer, A.H.; Vahala, G.
1989-08-01
During the past year we have studied stellarator equilibria with quasi-helical symmetry and the relation between the trajectories of the exact and the drift Hamiltonian. The relation between these trajectories is particularly important to issue of α particle confinement in a reactor. Work has also been done on the bootstrap current in the absence of symmetry, the effects of tearing modes on the current profile in a tokamak, and models of plasma turbulence. In addition, considerable time was spent during the year by Allen Boozer chairing the task force on Alternate Transport as part of the DoE transport initiative
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
Thermal stability of the tokamak plasma edge
International Nuclear Information System (INIS)
Stacey, W.M.
1997-01-01
The general linear, fluid, thermal instability theory for the plasma edge has been extended. An analysis of a two-dimensional fluid model of the plasma edge has identified the importance of many previously unappreciated phenomena associated with parallel and gyroviscous forces in the presence of large radial gradients, with large radial or parallel flows, with the temperature dependence of transport coefficients, and with the coupling of temperature, flow and density perturbations. The radiative condensation effect is generalized to include a further destabilizing condensation effect associated with radial heat conduction. Representative plasma edge neutral and impurity densities are found to be capable of driving thermal instabilities in the edge transport barrier and radiative mantle, respectively. (author)
Czech Academy of Sciences Publication Activity Database
Aftanas, Milan; Böhm, Petr; Bílková, Petra; Weinzettl, Vladimír; Zajac, Jaromír; Žáček, František; Stöckel, Jan; Hron, Martin; Pánek, Radomír; Scannell, R.; Walsh, M.
2012-01-01
Roč. 83, č. 10 (2012), 10E350-10E350 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] R&D Projects: GA ČR GA202/09/1467; GA MŠk 7G10072 Institutional research plan: CEZ:AV0Z20430508 Keywords : error analysis * Monte Carlo methods * plasma density * plasma diagnostics * plasma temperature * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://dx.doi.org/10.1063/1.4743956
Plasma Physics Network Newsletter, no. 5
1992-08-01
The fifth Plasma Physics Network Newsletter (IAEA, Vienna, Aug. 1992) includes the following topics: (1) the availability of a list of the members of the Third World Plasma Research Network (TWPRN); (2) the announcement of the fourteenth IAEA International Conference on Plasma Physics and Controlled Nuclear Fusion Research to be held in Wuerzburg, Germany, from 30 Sep. to 7 Oct. 1992; (3) the announcement of a Technical Committee Meeting on research using small tokamaks, organized by the IAEA as a satellite meeting to the aforementioned fusion conference; (4) IAEA Fellowships and Scientific Visits for the use of workers in developing member states, and for which plasma researchers are encouraged to apply through Dr. D. Banner, Head, Physics Section, IAEA, P.O. Box 100, A-1400 Vienna, Austria; (5) the initiation in 1993 of a new Coordinated Research Programme (CRP) on 'Development of Software for Numerical Simulation and Data Processing in Fusion Energy Research', as well as a proposed CRP on 'Fusion Research in Developing Countries using Middle- and Small-Scale Plasma Devices'; (6) support from the International Centre for Theoretical Physics (ICTP) for meetings held in Third World countries; (7) a report by W. Usada on Fusion Research in Indonesia; (8) News on ITER; (9) the Technical Committee Meeting planned 8-12 Sep. 1992, Canada, on Tokamak Plasma Biasing; (10) software made available for the study of tokamak transport; (11) the electronic mail address of the TWPRN; (12) the FAX, e-mail, and postal address for contributions to this plasma physics network newsletter.
Plasma physics network newsletter. No. 5
International Nuclear Information System (INIS)
1992-08-01
The fifth Plasma Physics Network Newsletter (IAEA, Vienna, August 1992) includes the following topics: (i) the availability of a list of the members of the Third World Plasma Research Network (TWPRN); (ii) the announcement of the fourteenth IAEA International Conference on Plasma Physics and Controlled Nuclear Fusion Research to be held in Wuerzburg, Germany, from September 30 to October 7, 1992; (iii) the announcement of a Technical Committee Meeting on research using small tokamaks, organized by the IAEA as a satellite meeting to the aforementioned fusion conference; (iv) IAEA Fellowships and Scientific Visits for the use of workers in developing member states, and for which plasma researchers are encouraged to apply through Dr. D. Banner, Head, Physics Section, IAEA, P.O. Box 100, A-1400 Vienna, Austria; (v) the initiation in 1993 of a new Coordinated Research Programme (CRP) on ''Development of Software for Numerical Simulation and Data Processing in Fusion Energy Research'', as well as a proposed CRP on ''Fusion Research in Developing Countries using Middle- and Small-Scale Plasma Devices''; (vi) support from the International Centre for Theoretical Physics (ICTP) for meetings held in Third World countries; (vii) a report by W. Usada on Fusion Research in Indonesia; (viii) News on ITER; (ix) the Technical Committee Meeting planned September 8-12, 1992, Canada, on Tokamak Plasma Biasing; (x) software made available for the study of tokamak transport; (xi) the electronic mail address of the TWPRN; (xii) and the FAX, e-mail and postal address for contributions to this plasma physics network newsletter (FAX: (43-1)-234564)
Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas
Energy Technology Data Exchange (ETDEWEB)
Lepson, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jernigan, J. Garrett [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2016-02-05
We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.
Investigation of plasma equilibrium in HL-1 tokamak
International Nuclear Information System (INIS)
Lu Zhihong; Jiang Yunxia; Yang Jinwei; Zhang Baozhu; Qiu Wei; Qin Yunwen
1987-01-01
In this paper, the plasma equilibrium in HL-1 tokamak has been discussed. The horizontal and vertical displacement of plasma is measured using a symmetical magneic probe system, and the temporal evolution of displacements is given by a data acquisition system with micro-computer. The influence of various stray fields on plasma equilibrium has been analysed. The direction and value of horizontal stray field induced by the totoidal field coils and primary windings are determined using vertical displacement data. By adjusting parameters of internal and external vertical field, in the flat part of discharge current, the plasma can be kept in its equilibuium state at the place where is 3 cm outer of chamber certre, i.e nearby the centre of limiter
The use of internal transport barriers in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Challis, C D [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)
2004-12-01
Internal transport barriers (ITBs) can provide high tokamak confinement at modest plasma current. This is desirable for operation with most of the current driven non-inductively by the bootstrap mechanism, as currently envisaged for steady-state power plants. Maintaining such plasmas in steady conditions with high plasma purity is challenging, however, due to MHD instabilities and impurity transport effects. Significant progress has been made in the control of ITB plasmas: the pressure profile has been varied using the barrier location; q-profile modification has been achieved with non-inductive current drive, and means have been found to affect density peaking and impurity accumulation. All these features are, to some extent, interdependent and must be integrated self-consistently to demonstrate a sound basis for extrapolation to future devices.
Positional instability analysis of tokamak plasmas by ERATO
International Nuclear Information System (INIS)
Kumagai, Michikazu; Tsunematsu, Toshihide; Tokuda, Shinji; Takeda, Tatsuoki
1983-06-01
The stability of axisymmetric modes of a tokamak plasma(positional instabilities) is analyzed for the Solov'ev equilibrium by using the linear ideal MHD code ERATO-J. The dependence of the stability on various parameters, i.e., the ellipticity and triangularity of the plasma cross-section, the aspect ratio, the safety factor at the magnetic axis, and the distance between the plasma and a conducting shell is investigated. Comparison of the results with those by the rigid model shows that the stability condition derived from the rigid model in terms of the decay index(n-index) of the external equilibrating field is a good approximation for the plasma with small triangular deformation. Also the results are compared with those of the rigid displacement model and applicability of the various models on the positional instability analyses is discussed. (author)
Plasma vertical instability in a tokamak with rail limiters
International Nuclear Information System (INIS)
Belashob, V.I.; Brevnov, N.N.; Gribov, Yu.V.; Putvinskij, S.V.
1989-01-01
An effect of currents between rail limiters on plasma equilibrium in the tokamak is studied theoretically and experimentally. Limiter currents can emerge at fast changes of plasma position along rail limiters for example when compression along major radius takes place and result in additional electrodynamic loadings on to the chamber and limiters. It is shown that at high currents between the limiters, the behaviour of discharge depends on limiter voltage polarity. When the plasma - limiter contact points are asymmetrically located respective to an equatorial plane a radial component of the limiter current emerges. The interaction of the component with the toroidal magnetic field can result in a vertical plasma filament instability. 9 refs.; 10 figs
Probabilistic analysis of tokamak plasma disruptions
International Nuclear Information System (INIS)
Sanzo, D.L.; Apostolakis, G.E.
1985-01-01
An approximate analytical solution to the heat conduction equations used in modeling component melting and vaporization resulting from plasma disruptions is presented. This solution is then used to propagate uncertainties in the input data characterizing disruptions, namely, energy density and disruption time, to obtain a probabilistic description of the output variables of interest, material melted and vaporized. (orig.)
10th International Conference and School on Plasma Physics and Controlled Fusion. Book of Abstracts
International Nuclear Information System (INIS)
Anon
2004-01-01
About 240 abstracts by Ukrainian and foreign authors submitted to 10-th International Conference and School on Plasma Physics and Controlled fusion have been considered by Conference Program Committee members. All the abstracts have been divided into 8 groups: magnetic confinement systems: stellarators, tokamaks, alternative conceptions; ITER and Fusion reactor aspects; basic plasma physics; space plasma; plasma dynamics and plasma-wall interaction; plasma electronics; low temperature plasma and plasma technologies; plasma diagnostics
International Nuclear Information System (INIS)
Dreicer, H.; Banton, M.E.; Ingraham, J.C.; Wittman, F.; Wright, B.L.
1976-01-01
The Experimental Plasma Physics group's main efforts continue to be directed toward the understanding of the mechanisms of electromagnetic energy absorption in a plasma, and the resultant plasma heating and energy transport. The high-frequency spectrum of plasma waves parametrically excited by the microwave signal at high powers has been measured. The absorption of a small test microwave signal in a plasma made parametrically unstable by a separate high-power driver microwave signal was also studied
Tokamak physics studies using x-ray diagnostic methods
International Nuclear Information System (INIS)
Hill, K.W.; Bitter, M.; von Goeler, S.
1987-03-01
X-ray diagnostic measurements have been used in a number of experiments to improve our understanding of important tokamak physics issues. The impurity content in TFTR plasmas, its sources and control have been clarified through soft x-ray pulse-height analysis (PHA) measurements. The dependence of intrinsic impurity concentrations and Z/sub eff/ on electron density, plasma current, limiter material and conditioning, and neutral-beam power have shown that the limiter is an important source of metal impurities. Neoclassical-like impurity peaking following hydrogen pellet injection into Alcator C and a strong effect of impurities on sawtooth behavior were demonstrated by x-ray imaging (XIS) measurements. Rapid inward motion of impurities and continuation of m = 1 activity following an internal disruption were demonstrated with XIS measurements on PLT using injected aluminum to enhance the signals. Ion temperatures up to 12 keV and a toroidal plasma rotation velocity up to 6 x 10 5 m/s have been measured by an x-ray crystal spectrometer (XCS) with up to 13 MW of 85-keV neutral-beam injection in TFTR. Precise wavelengths and relative intensities of x-ray lines in several helium-like ions and neon-like ions of silver have been measured in TFTR and PLT by the XCS. The data help to identify the important excitation processes predicted in atomic physics. Wavelengths of n = 3 to 2 silver lines of interest for x-ray lasers were measured, and precise instrument calibration techniques were developed. Electron thermal conductivity and sawtooth dynamics have been studied through XIS measurements on TFTR of heat-pulse propagation and compound sawteeth. A non-Maxwellian electron distribution function has been measured, and evidence of the Parail-Pogutse instability identified by hard x-ray PHA measurements on PLT during lower-hybrid current-drive experiments
Measurement of the effective plasma ion mass in large tokamaks
International Nuclear Information System (INIS)
Lister, J.B.; Villard, L.; Ridder, G. de
1997-01-01
There is not yet a straightforward method for the measurement of the D-T ratio in the centre of a tokamak plasma. One of the simpler measurements put forward in the past is the interpretation of the MHD spectrum in the frequency range of the Global Alfven Eigenmodes (GAE). However, the frequencies of these modes do not only depend on the plasma mass, but are also quite strongly dependent on the details of the current and density profiles, creating a problem of deconvolution of the estimate of the plasma mass from an implicit relationship between several measurable plasma parameters and the detected eigenmode frequencies. This method has been revised to assess its likely precision for the JET tokamak. The low n GAE modes are sometimes too close to the continuum edge to be detectable and the interpretation of the GAE spectrum is rendered less direct than had been hoped. We present a statistical study on the precision with which the D-T ratio could be estimated from the GAE spectrum on JET. (author) 4 figs., 8 refs
Magnetohydrodynamic equilibria and local stability of axisymmetric tokamak plasmas
International Nuclear Information System (INIS)
Peng, Y.K.M.; Dory, R.A.; Nelson, D.B.; Sayer, R.O.
1976-07-01
Axisymmetric magnetohydrodynamic equilibria are evaluated in terms of the Mercier Stability Criterion. The parameters of interest include poloidal beta (β/sub p/), current and pressure profile widths, D-shaped and doublet plasmas with elongation (sigma) and triangularity (delta), and the aspect ratio (A). For marginal local stability, the critical values of β, plasma current, and the safety factor q with fixed toroidal field at the geometric center of the plasma are obtained. It is shown that for a wide range of profiles in a D-shaped plasma with A = 3, the highest critical β occurs at β/sub p/ = 2.4, sigma = 1.65, and delta = 0.5. If the toroidal field at the coil surface is fixed, the highest critical pressure occurs near A approximately 3 to 4, given reasonable distance between the coils and the plasma edge. Calculations for a Doublet II-A plasma with sigma = 3 show that with similar pressure profile the highest critical β occurs at β/sub p/ = 1 and is 84 percent of the highest critical β for the D-shaped plasmas. Critical values of ohmic heating power density are also found to be comparable for the two plasma shapes. A D-shaped plasma with the above parameters is suggested for use in future high-β tokamak devices
Online Plasma Shape Reconstruction for EAST Tokamak
International Nuclear Information System (INIS)
Luo Zhengping; Xiao Bingjia; Zhu Yingfei; Yang Fei
2010-01-01
An online plasma shape reconstruction, based on the offline version of the EFIT code and MPI library, can be carried out between two adjacent shots in EAST. It combines online data acquisition, parallel calculation, and data storage together. The program on the master node of the cluster detects the termination of the discharge promptly, reads diagnostic data from the EAST mdsplus server on the completion of data storing, and writes the results onto the EFIT mdsplus server after the calculation is finished. These processes run automatically on a nine-nodes IBM blade center. The total time elapsed is about 1 second to several minutes, depending on the duration of the shot. With the results stored in the mdsplus server, it is convenient for operators and physicists to analyze the behavior of plasma using visualization tools.
International Nuclear Information System (INIS)
Ruskov, E.; Bell, M.; Budny, R.V.; McCune, D.C.; Medley, S.S.; Nazikian, R.; Synakowski, E.J.; Goeler, S. von; White, R.B.; Zweben, S.J.
1999-01-01
A case for substantial loss of fast ions degrading the performance of tokamak fusion test reactor plasmas [Phys. Plasmas 2, 2176 (1995)] with reversed magnetic shear (RS) is presented. The principal evidence is obtained from an experiment with short (40 - 70 ms) tritium beam pulses injected into deuterium beam heated RS plasmas [Phys. Rev. Lett. 82, 924 (1999)]. Modeling of this experiment indicates that up to 40% beam power is lost on a time scale much shorter than the beam - ion slowing down time. Critical parameters which connect modeling and experiment are: The total 14 MeV neutron emission, its radial profile, and the transverse stored energy. The fusion performance of some plasmas with internal transport barriers is further deteriorated by impurity accumulation in the plasma core. copyright 1999 American Institute of Physics
Physical domains in plasma physics
International Nuclear Information System (INIS)
Liboff, R.L.
1987-01-01
Do the plasma in the sun's core and the electron-conduction plasma in a semiconductor behave in the same way? This question is both fundamental and practical, for plasma physics plays a role in a vast area of natural phenomena and in many engineering devices. Understanding the cosmos, or designing a computer chip or a thermonuclear fusion reactor, requires first of all a realization of equations of motion that are appropriate to the particular problem. Similar physical differences occur in engineered structures. The plasmas in most thermonuclear fusion devices are basically like the plasma in the core of the sun: weakly coupled and classical - that is, obeying Newton's laws and Maxwell's equations. The conduction electrons in a semiconductor, on the other hand, obey the laws of quantum mechanics
Transient heat transport studies in JET conventional and advanced tokamak plasmas
International Nuclear Information System (INIS)
Mantica, P.; Coffey, I.; Dux, R.
2003-01-01
Transient transport studies are a valuable complement to steady-state analysis for the understanding of transport mechanisms and the validation of physics-based transport models. This paper presents results from transient heat transport experiments in JET and their modelling. Edge cold pulses and modulation of ICRH (in mode conversion scheme) have been used to provide detectable electron and ion temperature perturbations. The experiments have been performed in conventional L-mode plasmas or in Advanced Tokamak regimes, in the presence of an Internal Transport Barrier (ITB). In conventional plasmas, the issues of stiffness and non-locality have been addressed. Cold pulse propagation in ITB plasmas has provided useful insight into the physics of ITB formation. The use of edge perturbations for ITB triggering has been explored. Modelling of the experimental results has been performed using both empirical models and physics-based models. Results of cold pulse experiments in ITBs have also been compared with turbulence simulations. (author)
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.
International Nuclear Information System (INIS)
Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.
2001-01-01
The major increase in discharge duration and plasma energy in a next-step DT (deuterium-tritium) fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D (Research and Development) avenues for their resolution are presented
The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR
Energy Technology Data Exchange (ETDEWEB)
Tammen, H.F.
1995-01-10
One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics `Rijnhuizen`, was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL).
Energy Technology Data Exchange (ETDEWEB)
Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others
2001-01-10
The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.
The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR
International Nuclear Information System (INIS)
Tammen, H.F.
1995-01-01
One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics 'Rijnhuizen', was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL)
Plasma shape experiments for an optimized tokamak
International Nuclear Information System (INIS)
Hyatt, A.W.; Osborne, T.H.; Lazarus, E.A.
1994-07-01
In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, βτ E , in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed ''slot'' divertor geometry. power flux to the divertor floor using a closed ''slot'' divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes βτ E and divertor performance. The four shapes studied form a matrix of moderate and high elongations (κ congruent 1.8 and 2.1) and low and high triangularities (δ congruent 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot's transient normalized performance, β N H, where β N ≡ β/(I p )/aB T and H ≡ τ E /τ E ITER-89P , increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D
Plasma shape experiments for an optimized tokamak
Energy Technology Data Exchange (ETDEWEB)
Hyatt, A.W.; Osborne, T.H. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States)
1994-12-31
In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, {beta}{sub {tau}E}, in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed `slot` divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes {beta}{sub {tau}E} and divertor performance. The four shapes studied form a matrix of moderate and high elongations ({kappa} {approx_equal} 1.8 and 2.1) and low and high triangularities ({delta} {approx_equal} 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot`s transient normalized performance, {beta}{sub N}H, where {beta}{sub N} = {beta}/(I{sub p}/aB{sub T}) and H = {tau}{sub E}/{tau}{sub E}{sup ITER-89P}, increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D. (author) 7 refs., 7 figs.
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
Heavy Neutral Beam Probe for edge plasma analysis in tokamaks
International Nuclear Information System (INIS)
1991-01-01
The Heavy Neutral Beam Probe project presented in this document is part of an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadian de Fusion Magnetique. The overall objective of the effort is to apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes facility in Montreal, Canada. To achieve this goal, a research and development project was started in December, 1990 to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present, satisfactory progress has been achieved. The ion gun is fully operational with the neutralizer in the final assembly stage in preparation for testing. The beam diagnostics have been completed and mounted in the computer automated test stand. The analyzer design and detailed trajectory calculations are nearing completion to allow for the vacuum interface construction. The CAMAC based data acquisition system hardware was integrated into the test stand. Part of this hardware is a component of the Tokamak de Varennes' contribution to the collaboration. Next steps on the critical path include the beginning of the neutralization tests and the start of the analyzer construction. Anticipated installation of the diagnostic on the tokamak is Spring 1992
The evolution of the plasma current during tokamak disruptions
International Nuclear Information System (INIS)
Helander, P.; Andersson, F.; Anderson, D.; Lisak, M.; Eriksson, L.G.
2004-01-01
In a tokamak disruption, the ohmic plasma current is partly replaced by a current carried by runaway electrons. This process is analysed by combining the equations for runaway electron generation with Maxwell's equations for the evolution of the electric field. This allows a quantitative understanding to be gained of runaway production in present experiments, and extrapolation to be made to ITER. The runaway current typically becomes more peaked on the magnetic axis than the pre-disruption current. In fact, the central current density can rise although the total current falls, which may have implications for post-disruption plasma stability. Furthermore, it is found that the runaway current easily spreads radially in a filament way due to the high sensitivity of the runaway generation efficiency to plasma parameters. (authors)
Plasma rotation under a driven radial current in a tokamak
International Nuclear Information System (INIS)
Chang, C.S.
1999-01-01
The neoclassical behaviour of plasma rotation under a driven radial electrical current is studied in a tokamak geometry. An ambipolar radial electric field develops instantly in such a way that the driven current is balanced by a return current j p in the plasma. The j p x B torque pushes the plasma into a new rotation state both toroidally and poloidally. An anomalous toroidal viscosity is needed to avoid an extreme toroidal rotation speed. It is shown that the poloidal rotation relaxes to a new equilibrium speed, which is in general smaller than the E x B poloidal speed, and that the timescale for the relaxation of poloidal rotation is the same as that of toroidal rotation generation, which is usually much longer than the ion-ion collision time. (author)
Computations in plasma physics
International Nuclear Information System (INIS)
Cohen, B.I.; Killeen, J.
1984-01-01
A review of computer application in plasma physics is presented. Computer contribution to the investigation of magnetic and inertial confinement of a plasma and charged particle beam propagation is described. Typical utilization of computer for simulation and control of laboratory and cosmic experiments with a plasma and for data accumulation in these experiments is considered. Basic computational methods applied in plasma physics are discussed. Future trends of computer utilization in plasma reseaches are considered in terms of an increasing role of microprocessors and high-speed data plotters and the necessity of more powerful computer application
High power RF heating and nonthermal distributions in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Peeters, A.G.
1994-12-13
This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL).
High power RF heating and nonthermal distributions in tokamak plasmas
International Nuclear Information System (INIS)
Peeters, A.G.
1994-01-01
This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL)
Electromagnetic effects on trace impurity transport in tokamak plasmas
Hein, T.; Angioni, C.
2010-01-01
The impact of electromagnetic effects on the transport of light and heavy impurities in tokamak plasmas is investigated by means of an extensive set of linear gyrokinetic numerical calculations with the code GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and of analytical derivations with a fluid model. The impurity transport is studied by appropriately separating diffusive and convective contributions, and conditions of background microturbulence dominated by both ion temperature gradient (ITG) and trapped electron modes (TEMs) are analyzed. The dominant contribution from magnetic flutter transport turns out to be of pure convective type. However it remains small, below 10% with respect to the E ×B transport. A significant impact on the impurity transport due to an increase in the plasma normalized pressure parameter β is observed in the case of ITG modes, while for TEM the overall effect remains weak. In realistic conditions of high β plasmas in the high confinement (H-) mode with dominant ITG turbulence, the impurity diffusivity is found to decrease with increasing β in qualitative agreement with recent observations in tokamaks. In contrast, in these conditions, the ratio of the total off-diagonal convective velocity to the diagonal diffusivity is not strongly affected by an increase in β, particularly at low impurity charge, due to a compensation between the different off-diagonal contributions.
Ignition and burn control in tokamak plasmas
International Nuclear Information System (INIS)
Borrass, K.; Gruber, O.; Lackner, K.; Minardi, E.; Neuhauser, J.; Wilhelm, R.; Wunderlich, R.; Bromberg, L.; Cohn, D.R.
1981-01-01
Different schemes for the control of the thermal instability in an ignited fusion reactor are analysed by zero- and one-dimensional models. Passive stabilization methods considered are ripple-enhanced ion heat conduction, the effect of the major-radius variation of the plasma column in a time-independent vertical field, and the combination of both effects, including the spatial variation of the toroidal-ripple amplitude. Active control methods analysed are high-Q-driven operation and feedback-controlled major-radius variation following different scenarios. One-dimensional analyses taking into account only conductive losses show the existence of a single unstable mode in the energy balance, justifying, under these assumptions, the study of only global control. (author)
Study of edge turbulence in tokamak plasmas
International Nuclear Information System (INIS)
Sarazin, Y.
1997-01-01
The aim of this work is to propose a new frame to study turbulent transport in plasmas. In order to avoid the restraint of scale separability the forcing by flux is used. A critical one-dimension self-organized cellular model is developed. In keeping with experience the average transport can be described by means of diffusion and convection terms whereas the local transport could not. The instability due to interchanging process is thoroughly studied and some simplified equations are derived. The proposed model agrees with the following experimental results: the relative fluctuations of density are maximized on the edge, the profile shows an exponential behaviour and the amplitude of density fluctuations depends on ionization source strongly. (A.C.)
Controlled fusion and plasma physics
International Nuclear Information System (INIS)
Bickerton, R.J.
1991-01-01
On JET results were presented on additional heating power, on a recently discovered regime of enhanced pellet performance (PEP), on low-density H-mode plasma confinement with hot ions, bounds on very high electric currents by material limiters, the first experiments on lower hybrid current drive, on the L-H transition threshold dependence on the direction of the gradient-B drift, and on alpha-particle physics issues. The TFTR presentations focused on transport. Particle loss ramifications of the toroidal Alfven eigenmodes were found to be small, while their threshold of excitation is lower than theoretically predicted. On DIII-D a scaling study of transport with gyroradius as the only variable was reported, with approximately Bohm scaling emerging; but the effective heat diffusivity scaling could not be established due to profile consistency effects. While beta-limit investigations with DIII-D generally confirm the ideal, MHD limit found by Troyon, evidence of a reduction of the accessible range for the internal inductance with the safety factor seems to favour current-density control in a steady-state D-T burner. Onset of strongly sheared poloidal rotation in a thin layer during the L-H mode transition was experimentally shown, while a new, so-called VH (''very high'') confinement mode was discovered by boronization of the wall. The JT-90 tokamak has recently been upgraded to JT-60-U. Presentations by the ASDEX team summarized the lack of agreement with theory of L-mode confinement. With TEXTOR, an improved mode (I-mode) of confinement was found by boronization. Finally, reviews are included on the status of impurity transport and helium removal in tokamaks, on stellarators, alternative magnetic confinement systems, inertial confinement, and non-fusion plasma physics. 2 tabs
Magnetic fluctuations in the plasma of KT-5C tokamak
International Nuclear Information System (INIS)
Lu Ronghua; Pan Gesheng; Wang Zhijiang; Wen Yizhi; Yu Changxuan; Wan Shude; Liu Wandong; Wang Jun; Xu Min; Xiao Delong; Yu Yi
2004-01-01
A newly developed moveable magnetic probe array was installed on KT-5C tokamak. The profiles of radial and poloidal magnetic fluctuations of the plasma have been measured for (0.5r/a1.1). The experimental results indicate that there is a radial gradient which is greater than relative electrostatic fluctuations and the magnetic fluctuations contribute a little to losses. A strong coherence between fluctuations of 4 mm nearby two points suggests that the magnetic fluctuations have quite a long correlation length
Electron Heating of LHCD Plasma in HT-7 Tokamak
International Nuclear Information System (INIS)
Ding Yonghua; Wan Baonian; Lin Shiyao; Chen Zhongyong; Hu Xiwei; Shi Yuejiang; Hu Liqun; Kong Wei; Zhang Xiaoqing
2006-01-01
Electron heating via lower hybrid current drive (LHCD) has been investigated in HT-7 superconducting tokamak. Experiments show that the central electron temperature T e0 , the volume averaged electron temperature e > and the peaking factor of the electron temperature Q Te = T e0 / e > increase with the lower hybrid wave (LHW) power. Simultaneously the electron heating efficiency and the electron temperature as the function of the central line-averaged electron density (n e ) and the plasma current (I p ) have also been investigated. The experimental results are in a good agreement with those of the classical collision theory and the LHW power deposition theory
Self-consistent treatment of transport in tokamak plasmas
International Nuclear Information System (INIS)
Wilhelmsson, H.
1993-01-01
A theory is developed for the dynamics of tokamak plasmas considering the influence of combinations of simultaneous heating processes (alpha particle, auxiliary and ohmic), thermal conduction and particle diffusion, thermal and particle pinches, thermalization of alpha particles as well as the effects of boundary conditions. The analysis is based on a generalization of the central expansion technique which transforms the partial differential equations to a set of nonlinear coupled equations in time for the dynamic variables. Oscillatory solutions are found, but only in the presence of alpha particle heating. Examples of extensive computer simulations are included which support and complete the analytic results. (26 refs.)
Direction of Impurity Pinch and Auxiliary Heating in Tokamak Plasmas
International Nuclear Information System (INIS)
Angioni, C.; Peeters, A.G.
2006-01-01
A mechanism of particle pinch for trace impurities in tokamak plasmas, arising from the effect of parallel velocity fluctuations in the presence of a turbulent electrostatic potential, is identified analytically by means of a reduced fluid model and verified numerically with a gyrokinetic code for the first time. The direction of such a pinch reverses as a function of the direction of rotation of the turbulence in agreement with the impurity pinch reversal observed in some experiments when moving from dominant auxiliary ion heating to dominant auxiliary electron heating
Real-Time Software for the Compass Tokamak Plasma Control
Energy Technology Data Exchange (ETDEWEB)
Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Prague (Costa Rica)
2009-07-01
This poster presents the flexible and high-performance real time system that guarantees the desired time cycles for plasma control on the COMPASS tokamak: 500 {mu}s for toroidal field, current, equilibrium and shaping; 50 {mu}s for fast control of the equilibrium and vertical instability. This system was developed on top of a high-performance processor and a software framework (MARTe) tailored for real-time. The preliminary measurements indicate that the time constraints will be met on the final solution. The system allows the making of modifications in the future to improve software components. (A.C.)
A theory of the coherent fundamental plasma emission in Tokamaks
International Nuclear Information System (INIS)
Alves, M.V.; Chian, A.C.-L.
1987-01-01
A theoretical model of coherent radiation near the fundamental plasma frequency in tokamaks is proposed. It is shown that, in the presence of runaway electrons, the beam-generated Langmuir waves (L) can be parametrically converted into electromagnetic waves (T) through ponderomotive coupling to ion acoustic waves (S). Two types of pumps are considered: travelling wave pump and standing wave pump. Expressions are derived for the excitation conditions and the growth rates of electromagnetic decay instabilities (L-> T + S), electromagnetic fusion instabilities (L + S -> T) and electromagnetic oscillating two-stream instabilities (L -> T+- S * , where S * is a purely growing mode). (author) [pt
Information content of transient synchrotron radiation in tokamak plasmas
International Nuclear Information System (INIS)
Fisch, N.J.; Kritz, A.H.
1989-04-01
A brief, deliberate, perturbation of hot tokamak electrons produces a transient, synchrotron radiation signal, in frequency-time space, with impressive informative potential on plasma parameters; for example, the dc toroidal electric field, not available by other means, may be measurably. Very fast algorithms have been developed, making tractable a statistical analysis that compares essentially all parameter sets that might possibly explain the transient signal. By simulating data numerically, we can estimate the informative worth of data prior to obtaining it. 20 refs., 2 figs
A theory of the coherent fundamental plasma emission in Tokamaks
International Nuclear Information System (INIS)
Alves, M.V.; Chian, A.C.-L.
1987-07-01
A theoretical model of coherent radiation near the fundamental plasma frequency in Tokamaks is proposed. It is shown that, in the presence of runaway electrons, the beam-generated Langmuir waves (L) can be paarmetrically converted into electromagnetic waves (T) through ponderomotive coupling to ion acoustic waves (S). Two types of pumps are considered: traveling wave and standing wave pump. Expressions are derived for the excitation conditions and the growth rates of electomagnetic decay instabilities (L → T + S), electromagnetic fusion instabilities (L + S → T) and electromagnetic oscillating two-stream instabilities (L → T+-S sup(*) is a purely growing mode). (author) [pt
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
Fusion programs in Applied Plasma Physics
International Nuclear Information System (INIS)
1992-07-01
The Applied Plasma Physics (APP) program at General Atomics (GA) described here includes four major elements: (a) Applied Plasma Physics Theory Program, (b) Alpha Particle Diagnostic, (c) Edge and Current Density Diagnostic, and (d) Fusion User Service Center (USC). The objective of the APP theoretical plasma physics research at GA is to support the DIII-D and other tokamak experiments and to significantly advance our ability to design a commercially-attractive fusion reactor. We categorize our efforts in three areas: magnetohydrodynamic (MHD) equilibria and stability; plasma transport with emphasis on H-mode, divertor, and boundary physics; and radio frequency (rf). The objective of the APP alpha particle diagnostic is to develop diagnostics of fast confined alpha particles using the interactions with the ablation cloud surrounding injected pellets and to develop diagnostic systems for reacting and ignited plasmas. The objective of the APP edge and current density diagnostic is to first develop a lithium beam diagnostic system for edge fluctuation studies on the Texas Experimental Tokamak (TEXT). The objective of the Fusion USC is to continue to provide maintenance and programming support to computer users in the GA fusion community. The detailed progress of each separate program covered in this report period is described in the following sections
Fourth Latin-American workshop on plasma physics. Contributed papers
International Nuclear Information System (INIS)
1990-01-01
The main goal of this series of Workshops is to provide a periodic meeting place for Latin-American researchers in plasma physics together with colleagues from other countries around the world. This volume includes the contributed papers presented at the Workshop on Plasma Physics held in Buenos Aires in 1990. The scope of the Workshop can be synthesized in the following main subjects: Tokamak experiments and theory; alternative confinement systems and basic experiments; technology and applications; general theory; astrophysical and space plasmas
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
International Nuclear Information System (INIS)
Takase, Haruhiko; Senda, Ikuo
1999-01-01
A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)
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
International Nuclear Information System (INIS)
Stefanovskii, A. M.
2011-01-01
The processes that are likely to accompany discharge disruptions and sawteeth in a tokamak are considered in a simple plasma current model. The redistribution of the current density in plasma is supposed to be primarily governed by the onset of the MHD-instability-driven turbulent plasma mixing in a finite region of the current column. For different disruption conditions, the variation in the total plasma current (the appearance of a characteristic spike) is also calculated. It is found that the numerical shape and amplitude of the total current spikes during disruptions approximately coincide with those measured in some tokamak experiments. Under the assumptions adopted in the model, the physical mechanism for the formation of the spikes is determined. The mechanism is attributed to the diffusion of the negative current density at the column edge into the zero-conductivity region. The numerical current density distributions in the plasma during the sawteeth differ from the literature data.
Theoretical plasma physics. Final report
International Nuclear Information System (INIS)
Vahala, G.; Tracy, E.
1996-04-01
During the past year, the authors have concentrated on (1) divertor physics, (2) thermo-lattice Boltzmann (TLBE) approach to turbulence, and (3) phase space techniques in gyro-resonance problems in collaboration with Dieter Sigmar (MIT), Sergei Krasheninnikov (MIT), Linda Vahala (ODU), Joseph Morrison (AS and M/NASA-Langley), Pavol Pavlo and Josef Preinhaelter (institute of Plasma Physics, Czech Academy of Sciences) and Allan Kaufman (LBL/U.C.Berkeley). Using a 2-equation compressible closure model with a 2D mean flow, the authors are investigating the effects of 3D neutral turbulence on reducing the heat load to the divertor plate by various toroidal cavity geometries. These studies are being extended to examine 3D mean flows. Thermal Lattice Boltzmann (TLBE) methods are being investigated to handle 3D turbulent flows in nontrivial geometries. It is planned to couple the TLBE collisional regime to the weakly collisional regime and so be able to tackle divertor physics. In the application of phase space techniques to minority-ion RF heating, resonance heating is treated as a multi-stage process. A generalization of the Case-van Kampen analysis is presented for multi-dimensional non-uniform plasmas. Effects such as particle trapping and the ray propagation dynamics in tokamak geometry can now be handled using Weyl calculus
Plasma physics for controlled fusion
Miyamoto, Kenro
2016-01-01
This new edition presents the essential theoretical and analytical methods needed to understand the recent fusion research of tokamak and alternate approaches. The author describes magnetohydrodynamic and kinetic theories of cold and hot plasmas in detail. The book covers new important topics for fusion studies such as plasma transport by drift turbulence, which depend on the magnetic configuration and zonal flows. These are universal phenomena of microturbulence. They can modify the onset criterion for turbulent transport, instabilities driven by energetic particles as well as alpha particle generation and typical plasma models for computer simulation. The fusion research of tokamaks with various new versions of H modes are explained. The design concept of ITER, the international tokamak experimental reactor, is described for inductively driven operations as well as steady-state operations using non-inductive drives. Alternative approaches of reversed-field pinch and its relaxation process, stellator includi...
Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor
International Nuclear Information System (INIS)
Bell, M.G.; Beer, M.
1997-02-01
Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l i ). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q a ∼ 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l i plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q 0 > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions
New DIII-D tokamak plasma control system
International Nuclear Information System (INIS)
Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.
1992-09-01
A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter
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.)
Electron and current density measurements on tokamak plasmas
International Nuclear Information System (INIS)
Lammeren, A.C.A.P. van.
1991-01-01
The first part of this thesis describes the Thomson-scattering diagnostic as it was present at the TORTUR tokamak. For the first time with this diagnostic a complete tangential scattering spectrum was recorded during one single laser pulse. From this scattering spectrum the local current density was derived. Small deviations from the expected gaussian scattering spectrum were observed indicating the non-Maxwellian character of the electron-velocity distribution. The second part of this thesis describes the multi-channel interferometer/ polarimeter diagnostic which was constructed, build and operated on the Rijnhuizen Tokamak Project (RTP) tokamak. The diagnostic was operated routinely, yielding the development of the density profiles for every discharge. When ECRH (Electron Cyclotron Resonance Heating) is switched on the density profile broadens, the central density decreases and the total density increases, the opposite takes place when ECRH is switched off. The influence of MHD (magnetohydrodynamics) activity on the density was clearly observable. In the central region of the plasma it was measured that in hydrogen discharges the so-called sawtooth collapse is preceded by an m=1 instability which grows rapidly. An increase in radius of this m=1 mode of 1.5 cm just before the crash is observed. In hydrogen discharges the sawtooth induced density pulse shows an asymmetry for the high- and low-field side propagation. This asymmetry disappeared for helium discharges. From the location of the maximum density variations during an m=2 mode the position of the q=2 surface is derived. The density profiles are measured during the energy quench phase of a plasma disruption. A fast flattening and broadening of the density profile is observed. (author). 95 refs.; 66 figs.; 7 tabs
International Nuclear Information System (INIS)
Anon.
1979-01-01
Applied Plasma Physics is a major sub-organizational unit of the Magnetic Fusion Energy (MFE) Program. It includes Fusion Plasma Theory and Experimental Plasma Research. The Fusion Plasma Theory group has the responsibility for developing theoretical-computational models in the general areas of plasma properties, equilibrium, stability, transport, and atomic physics. This group has responsibility for giving guidance to the mirror experimental program. There is a formal division of the group into theory and computational; however, in this report the efforts of the two areas are not separated since many projects have contributions from members of both. Under the Experimental Plasma Research Program we are developing a neutral-beam source, the intense, pulsed ion-neutral source (IPINS), for the generation of a reversed-field configuration on 2XIIB. We are also studying the feasibility of using certain neutron-detection techniques as plasma diagnostics in the next generation of thermonuclear experiments
International Nuclear Information System (INIS)
Anon.
1978-01-01
Applied Plasma Physics is a major sub-organizational unit of the MFE Program. It includes Fusion Plasma Theory and Experimental Plasma Research. The Fusion Plasma Theory group has the responsibility for developing theoretical-computational models in the general areas of plasma properties, equilibrium, stability, transport, and atomic physics. This group has responsibility for giving guidance to the mirror experimental program. There is a formal division of the group into theory and computational; however, in this report the efforts of the two areas are not separated since many projects have contributions from members of both. Under the Experimental Plasma Research Program, we are developing the intense, pulsed neutral-beam source (IPINS) for the generation of a reversed-field configuration on 2XIIB. We are also studying the feasibility of utilizing certain neutron-detection techniques as plasma diagnostics in the next generation of thermonuclear experiments
Princeton University, Plasma Physics Laboratory annual report, October 1, 1988--September 30, 1989
Energy Technology Data Exchange (ETDEWEB)
1989-12-31
This report contains discussions on the following topics: principal parameters achieved in experimental devices (FY89); tokamak fusion test reactor; compact ignition tokamak; princeton beta experiment- modification; current drive experiment; international collaboration; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical studies; tokamak modeling; international thermonuclear experimental reactor; engineering department; project planning and safety office; quality assurance and reliability; technology transfer; administrative operations; PPPL patent invention disclosures for (FY89); graduate education: plasma physics; graduate education: plasma science and technology; and Princeton Plasmas Physics Laboratory Reports (FY89).
Princeton Plasma Physics Laboratory. Annual report, October 1, 1989--September 30, 1990
Energy Technology Data Exchange (ETDEWEB)
1990-12-31
This report discusses the following topics: principal parameters achieved in experimental devices fiscal year 1990; tokamak fusion test reactor; compact ignition tokamak; Princeton beta experiment- modification; current drive experiment-upgrade; international collaboration; x-ray laser studies; spacecraft glow experiment; plasma processing: deposition and etching of thin films; theoretical studies; tokamak modeling; international thermonuclear experimental reactor; engineering department; project planning and safety office; quality assurance and reliability; technology transfer; administrative operations; PPPL patent invention disclosures for fiscal year 1990; graduate education; plasma physics; graduate education: plasma science and technology; science education program; and Princeton Plasma Physics Laboratory reports fiscal year 1990.
Princeton University, Plasma Physics Laboratory annual report, October 1, 1988--September 30, 1989
Energy Technology Data Exchange (ETDEWEB)
1989-01-01
This report contains discussions on the following topics: principal parameters achieved in experimental devices (FY89); tokamak fusion test reactor; compact ignition tokamak; princeton beta experiment- modification; current drive experiment; international collaboration; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical studies; tokamak modeling; international thermonuclear experimental reactor; engineering department; project planning and safety office; quality assurance and reliability; technology transfer; administrative operations; PPPL patent invention disclosures for (FY89); graduate education: plasma physics; graduate education: plasma science and technology; and Princeton Plasmas Physics Laboratory Reports (FY89).
Princeton University, Plasma Physics Laboratory annual report, October 1, 1988--September 30, 1989
International Nuclear Information System (INIS)
1989-01-01
This report contains discussions on the following topics: principal parameters achieved in experimental devices (FY89); tokamak fusion test reactor; compact ignition tokamak; princeton beta experiment- modification; current drive experiment; international collaboration; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical studies; tokamak modeling; international thermonuclear experimental reactor; engineering department; project planning and safety office; quality assurance and reliability; technology transfer; administrative operations; PPPL patent invention disclosures for (FY89); graduate education: plasma physics; graduate education: plasma science and technology; and Princeton Plasmas Physics Laboratory Reports (FY89)
Static and dynamic control of plasma equilibrium in a Tokamak
International Nuclear Information System (INIS)
Blum, J.; Dei Cas, R.
1979-01-01
We are dealing here with the problem of controlling the plasma boundary and its displacements. Static control consists in determining the currents in the external coils of the Tokamak so that the plasma boundary has certain fixed characteristics: radial position, vertical elongation, desired shape. A self-consistent method is proposed here, considering a free plasma boundary, and using the techniques of optimal control of distributed parameter systems to solve the problem. The dynamic control problem considered in the second part of the paper is the control of the plasma radial displacements. An elaborate system of preprogramming and feedback control has been developed to ensure equilibrium and stability of the horizontal plasma motions. Optimal control techniques have been used to calculate the optimal primary coils configuration, the preprogramming voltages and the feedback gains. A new stability diagrams has been obtained which takes into account the erosion of the plasma by the limiter. All these calculations have been applied successfully to TFR 600 where thin liner and the presence of an iron core make the problem of stabilization of the radial displacements very difficult
Plasma confinement using biased electrode in the TCABR tokamak
International Nuclear Information System (INIS)
Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.; Machida, M.
2005-01-01
Experimental data obtained on the TCABR tokamak (R = 0.61 m, r = 0.18 m) with an electrally polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1), and hydrogen gas injection adjusted for keeping the electron density at 1.0x10(19) m(-3) without bias. Temporal and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the density, up to a maximum factor of 2.6, while H-alpha hydrogen spectral line intensity decreases, and the CIII impurity stays on the same level. The analysis of temporal and radial profiles of plasma parameters indicates that the confined plasma entered in the H-mode regime. The data analysis shows a maximum enhanced confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the plasma edge, using measured floating potentials and ion saturation currents, show strong decrease in the power spectra and transport. Bifurcation was not observed, and the decrease in the saturation current occurs in 50 microseconds. (author)
Heavy Neutral Beam Probe for edge plasma analysis in Tokamaks
International Nuclear Information System (INIS)
Castracane, J.; Saravia, E.; Beckstead, J.; Aceto, S.
1993-01-01
The contents of this report present the progress achieved to date on the Heavy Neutral Beam Probe project. This effort is an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadien de Fusion Magnetique (CCFM). The overall objective of the effort is to develop and apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes (TdeV) facility in Montreal, Canada. To achieve this goal, a research and development project was established to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present the project is in the middle of its second budget period with the instrumentation on-site at TdeV. The first half of this budget period was used to complete total system tests at InterScience, Inc., dismantle and ship the hardware to TdeV, re-assemble and install the HNBP on the tokamak. Integration of the diagnostic into the TdeV facility has progressed to the point of first beam production and measurement on the plasma. At this time, the HNBP system is undergoing final de-bugging prior to re-start of machine operation in early Fall of this year
Anomalous energy transport in hot plasmas: solar corona and Tokamak
International Nuclear Information System (INIS)
Beaufume, P.
1992-04-01
Anomalous energy transport is studied in two hot plasmas and appears to be associated with a heating of the solar corona and with a plasma deconfining process in tokamaks. The magnetic structure is shown to play a fundamental role in this phenomenon through small scale instabilities which are modelized by means of a nonlinear dynamical system: the Beasts' Model. Four behavior classes are found for this system, which are automatically classified in the parameter space thanks to a neural network. We use a compilation of experimental results relative to the solar corona to discuss current-based heating processes. We find that a simple Joule effect cannot provide the required heating rates, and therefore propose a dimensional model involving a resistive reconnective instability which leads to an efficient and discontinuous heating mechanism. Results are in good agreement with the observations. We give an analytical expression for a diffusion coefficient in tokamaks when magnetic turbulence is perturbing the topology, which we validate thanks to the standard mapping. A realistic version of the Beasts' Model allows to test a candidate to anomalous transport: the thermal filamentation instability
Internal transport barrier physics for steady state operation in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Wakatani, Masahiro [Kyoto Univ., Graduate School of Engineering, Uji, Kyoto (Japan); Fukuda, Takeshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Connor, Jack W. [Culham Science Centre, EURATOM/UKAEA Association (United Kingdom); Garbet, Xavier [Culham Science Centre, EFDA-JET CSU (United Kingdom); Gormezano, Claude [Associazone EURATOM-ENEA sulla Fusione C.R. Frascati (Italy); Mukhovatov, Vladimir [ITER Naka Joint Work Site, ITER Physics Unit, Naka, Ibaraki (Japan)
2003-07-01
Experimental results for the ITB (Internal Transport Barrier) formation and sustainment are compiled in a unified manner to find common features of ITBs in tokamaks. Global scaling laws for threshold power to obtain the ITBs are discussed. Theoretical models for plasmas with ITBs are summarized from stability and transport point of view. Finally possibility to obtain steady-state ITBs will be discussed in addition to extrapolation to ITER. (author)
Langmuir probe evaluation of the plasma potential in tokamak edge plasma for non-Maxwellian EEDF
Energy Technology Data Exchange (ETDEWEB)
Popov, Ts.K. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Dimitrova, M. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Ivanova, P. [Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Hasan, E. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Horacek, J.; Dejarnac, R.; Stoeckel, J.; Weinzettl, V. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Kovacic, J. [Jozef Stefan Institute, Ljubljana (Slovenia)
2014-04-15
The First derivative probe technique for a correct evaluation of the plasma potential in the case of non-Maxwellian EEDF is presented and used to process experimental data from COMPASS tokamak. Results obtained from classical and first derivative techniques are compared and discussed. The first derivative probe technique provides values for the plasma potential in the scrape-off layer of tokamak plasmas with an accuracy of about ±10%. Classical probe technique can provide values of the plasma potential only, if the electron and ion temperatures are known as well as the coefficient of secondary electron emission. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
Kono, Mitsuo
2010-01-01
A nonlinearity is one of the most important notions in modern physics. A plasma is rich in nonlinearities and provides a variety of behaviors inherent to instabilities, coherent wave structures and turbulence. The book covers the basic concepts and mathematical methods, necessary to comprehend nonlinear problems widely encountered in contemporary plasmas, but also in other fields of physics and current research on self-organized structures and magnetized plasma turbulence. The analyses make use of strongly nonlinear models solved by analytical techniques backed by extensive simulations and available experiments. The text is written for senior undergraduates, graduate students, lecturers and researchers in laboratory, space and fusion plasmas.
Poloidal asymmetries in the limiter shadow plasma of the Alcator C tokamak. Volume 1
International Nuclear Information System (INIS)
LaBombard, B.
1986-05-01
This thesis investigates conditions which exist in the limiter shadow plasma of the Alcator C tokamak. The understanding of this edge plasma region is approached from both experimental and theoretical points of view. First, a general overview of edge plasma physical processes is presented. Simple edge plasma models and conditions which can theoretically result in a poloidally asymmetric edge plasma are discussed. A review of data obtained from previous diagnostics in the Alcator C edge plasma is then used to motivate the development of a new edge plasma diagnostic system (DENSEPACK) to experimentally investigate poloidal asymmetries in this region. The bulk of this thesis focuses on the marked poloidal asymmetries detected by this poloidal probe array and possible mechanisms which might support such asymmetries on a magnetic flux surface. In processing the probe data, some important considerations on fitting Langmuir probe characteristics are identified. The remainder of this thesis catalogues edge versus central plasma parameter dependences. Regression analysis techniques are applied to characterize edge density for various central plasma parameters. Edge plasma conditions during lower hybrid radio frequency heating and pellet injection are also discussed
Two dimensional neutral transport analysis in tokamak plasma
International Nuclear Information System (INIS)
Shimizu, Katsuhiro; Azumi, Masafumi
1987-02-01
Neutral particle influences the particle and energy balance, and play an important role on sputtering impurity and the charge exchange loss of neutral beam injection. In order to study neutral particle behaviour including the effects of asymmetric source and divertor configuration, the two dimensional neutral transport code has been developed using the Monte-Carlo techniques. This code includes the calculation of the H α radiation intensity based on the collisional-radiation model. The particle confinement time of the joule heated plasma in JT-60 tokamak is evaluated by comparing the calculated H α radiation intensity with the experimental data. The effect of the equilibrium on the neutral density profile in high-β plasma is also investigated. (author)
Remote network control plasma diagnostic system for Tokamak T-10
International Nuclear Information System (INIS)
Troynov, V I; Zimin, A M; Krupin, V A; Notkin, G E; Nurgaliev, M R
2016-01-01
The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet. (paper)
Heat flow during sawtooth collapse in tokamak plasmas
International Nuclear Information System (INIS)
Hanada, Kazuaki
1994-01-01
Heat flow during sawtooth collapse was studied on the WT-3 tokamak by using temporal evolution of soft X-ray intensity profile in the poloidal cross section in a lower hybrid current driven plasma as well as an electron cyclotron heated plasma. Two phase in sawtooth collapses were observed. In the first phases, the hottest spot that is the peak of the soft X-ray distribution approaches the inversion surface and heat flows out through a narrow gate on the inversion surface. In the second phase, the hottest spot stays on the inversion surface, and heat flows out through the whole inversion surface. This suggests that magnetic reconnection as predicted by Kadomtsev's model occurs in the first phase, but in the second phase, a different mechanism dominates heat flow. (author)
LH-power coupling in advanced tokamak plasmas in JET
International Nuclear Information System (INIS)
Joffrin, E.; Erents, K.; Gormezano, C.
2000-02-01
Lower Hybrid Current Drive (LHCD) is the most efficient tool to generate non-inductive current in tokamak plasmas. In JET, significant modifications of the current profile have been recently achieved in coupling up to 3MW of LH power in optimised shear discharges. However, the improved particle confinement during optimised shear plasmas results in a sharp decrease of the electron density in front the launcher close or below the cut-off density (ne=1.7.10 17 m -3 for f LH =37GHz) and makes difficult the coupling of the LH power. Deuterium gas near the launcher can help to improve the coupling, but has also the effect of increasing the ELM activity leading to the erosion of the internal transport barrier (ITB). Future development of lower hybrid launcher should include the constraints imposed by scenario such as the optimised shear. (author)
Study of the plasma edge turbulence in tokamaks
International Nuclear Information System (INIS)
Garbet, X.; Laurent, L.; Mourgues, F.; Roubin, J.P.; Samain, A.
1990-01-01
The plasma edge in tokamaks is known to be very turbulent. We investigate here the non linear stability of a test mode in presence of an helical potential perturbation, i.e. a pump mode, which simulates the plasma turbulence. The particle trajectories in this perturbed equilibrium are derived using an hamiltonian formalism. The electrons appear to have trapped trajectories in the potential well of the pump mode, while the ions experience a large convective motion. These two effects have a large influence on the test mode stability. First, non linearly trapped electrons supply an energy source for the test mode. Second, the ion convective motion introduces a radial scale of the test mode larger than the ion Larmor radius, in agreement with experimental data. These two phenomena allow a bifurcation in the turbulence level and provide therefore an explanation for the L-H transition
Integrated Tokamak modeling: When physics informs engineering and research planning
Poli, Francesca Maria
2018-05-01
Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.
International Nuclear Information System (INIS)
2003-01-01
Theoretical and experimental short communications are presented on plasma and fusion covering the following subjects: plasma production, confinement, plasma waves, diagnostics, heating, tokamak, impurities, astrophysics plasma and technological applications
High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks
International Nuclear Information System (INIS)
Goodall, D.H.J.
1982-01-01
High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)
12th Czechoslovak seminar on plasma physics and technology
International Nuclear Information System (INIS)
1983-03-01
The 12th Czechoslovak seminar on plasma physics and technology was oriented mainly to the problems of high-temperature plasmas and controlled thermonuclear fusion. The proceedings contain 27 invited papers and communications presented in three sections: 1) Inertial controlled fusion, 2) Tokamaks, 3) Theory and miscellaneous topics. The first group of papers deals with various problems of electron-beam, ion-beam, and laser fusion, including physical processes in fusion targets. The tokamak section discusses the latest experimental results achieved in the Russian tokamaks FT-2, Tuman 2-a, T-7 and T-10, in the Czechoslovak tokamak TM-1-MH, and in the Hungarian tokamak MT-1. A detailed survey is presented of work on neutral atom injectors in Novosibirsk. In the third section several papers on theoretical studies of nonlinear and turbulent processes in a hot plasma are presented together with a simulation study of a hybrid tokamak reactor. Several contributions on special diagnostic methods are presented. (J.U.)
Plasma confinement using biased electrode in the TCABR tokamak
International Nuclear Information System (INIS)
Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Machida, M.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.
2005-01-01
Experimental data obtained on the TCABR tokamak (R = 0.61 m, a = 0.18 m) with an electrically polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1) and hydrogen gas injection adjusted for keeping the electron density at 1.0 x 10 19 m -3 without bias. Time evolution and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the central line-averaged density, up to a maximum factor of 2.6, while H α hydrogen spectral line intensity decreases and the C III impurity stays on the same level. The analysis of temporal behaviour and radial profiles of plasma parameters indicates that the confined plasma enters the H-mode regime. The data analysis shows a maximum enhanced energy confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the Scrape-Off-Layer, using measured floating potentials and ion saturation currents, show a strong decrease in the power spectra and transport. Bifurcation was not observed and the decrease in the saturation current occurs in 50 μs
Radio frequency plasma heating in large tokamak systems near the lower hybrid resonance
International Nuclear Information System (INIS)
Deitz, A.; Hooke, W.M.
1975-01-01
The frequency range, power, efficiency, and pulse length of a high power rf system are discussed as they might be applied to the TFTR Tokamak facility as well as on a full scale reactor. Comparisons are made of the size, power output, and costs to obtain microwave power sufficient to satisfy the physics requirements. A new microwave feed concept is discussed which will improve the coupling of the microwave energy into the plasma. The unique advantages of waveguide feed systems is apparent when one considers the practical problems associated with coupling supplementary heating energy into a reactor
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)
Dynamics and feedback control of plasma equilibrium position in a tokamak
International Nuclear Information System (INIS)
Burenko, O.
1983-01-01
A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems
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
The conversion of the thermal energy of plasma in the SOL of tokamaks
International Nuclear Information System (INIS)
Nedospasov, A.V.; Nenova, N.V.
2008-01-01
When the plasma expands across the confining magnetic field, a part of its thermal energy is converted to electrical energy. In the SOL of tokamaks, the motion of the plasma across the field due to turbulent processes is accompanied by its departure along the open lines of the magnetic field. The conversion of thermal energy is taken into account in theoretical studies devoted to the physics of plasma in the SOL; however, this conversion is ignored in numerical models, for example, in B2-SOLPS4.0. This paper deals with thermal-to-electrical energy conversion in the SOL of tokamaks. It is demonstrated that the part of the thermal energy subjected to conversion to electrical energy forms an appreciable part of the total energy flowing in the SOL. In ITER, this fraction may be as high as 20-25%. The electrical energy generated in the SOL volume is liberated in the form of Joule heat in a relatively cold plasma in the vicinity of diverter plates or directly on these plates. (letter)
Helical temperature perturbations associated with tearing modes in tokamak plasmas
International Nuclear Information System (INIS)
Fitzpatrick, R.
1994-06-01
An investigation is made into the electron temperature perturbations associated with tearing modes in tokamak plasmas, with a view to determining the mode structure using Electron Cyclotron Emission (ECE) data. It is found that there is a critical magnetic island width below which the conventional picture where the temperature is flattened inside the separatrix is invalid. This effect comes about because of the stagnation of magnetic field lines in the vicinity of the rational surface and the finite parallel thermal conductivity of the plasma. For islands whose widths lie below the critical value there is no flattening of the electron temperature inside the separatrix. Such islands have quite different ECE signatures to conventional magnetic islands. In fact the two island types could, in principle, be differentiated experimentally. It should also be possible to map out the outer ideal magnetohydrodynamical eigenfunctions using ECE data. Islands whose widths are much less than the critical value are not destabilized by the perturbed bootstrap current, unlike conventional magnetic islands. This effect is found to have a number of very interesting consequences and may, indeed, provide an explanation for some puzzling experimental results regarding error field induced magnetic reconnection. All islands whose widths are much greater than the critical width possess a boundary layer on the separatrix which enables heat to be transported from one side of the island to the other via the X-point region. The structure of this boundary layer is described in some detail. Finally, the critical island width is found to be fairly substantial in conventional tokamak plasmas, provided that the long mean free path nature of parallel heat transport and the anomalous nature of perpendicular heat transport are taken into account in the calculation
Formation and sustainment of a low aspect ratio tokamak by a series of plasma injections
International Nuclear Information System (INIS)
Shimamura, Shin; Taniguchi, Makoto; Takahashi, Tsutomu; Nogi, Yasuyuki
1995-01-01
A low aspect ratio tokamak plasma was generated and sustained by injecting a series of plasmas from a magnetized coaxial gun into a flux conserver with toroidal field. The magnetized coaxial gun was supplied by an oscillating current with a d.c. component. The first few current pulses injected plasma and helicity into the flux conserver. This pulse helicity injection method worked effectively to maintain the low aspect ratio tokamak. 8 refs., 5 figs
International Nuclear Information System (INIS)
Mazzucato, E.
2000-01-01
The next step in the demonstration of the scientific feasibility of a tokamak fusion reactor is a DT burning plasma experiment for the study and control of self-heated plasmas. In this paper, the authors examine the role of the toroidal magnetic field on the confinement of a tokamak plasma in the ELMy H-mode regime--the operational regime foreseen for ITER
The role of drifts in the plasma transport at the tokamak core–SOL interface
Energy Technology Data Exchange (ETDEWEB)
Chankin, A.V., E-mail: Alex.Chankin@ipp.mpg.de [Max-Planck-Institut für Pasmaphysik, EURATOM Association, 85748 Garching (Germany); Coster, D.P. [Max-Planck-Institut für Pasmaphysik, EURATOM Association, 85748 Garching (Germany)
2013-07-15
The interface between the core (inside the magnetic separatrix in X-point configurations) and the scrape-off layer (SOL) of tokamaks is a delicate region of the magnetic topology transition from closed to open field lines where neither the standard neoclassical theory nor the SOL physics fully apply. Sharp gradients of plasma parameters in the outer core, caused by the proximity of divertor sinks in the near SOL, invalidate some ordering assumptions of the neoclassical theory. At the same time, the existence of closed flux surfaces in the core enforces ambipolarity of radial plasma flows, in difference to the situation in the SOL where the current loop may close through the divertor. Detailed analysis of the plasma transport and flows with the emphasis on the outer core region, just inside the separatrix, is carried out in the paper, based on EDGE2D modelling and analytical formulas.
International Nuclear Information System (INIS)
Conn, R.W.; Mau, T.K.; Prinja, A.K.
1983-01-01
A physical model for the space and time evolution of the primary parameters of ordinary and burning tokamak plasmas is described by employing a fluid plasma treatment coupled to a magnetohydrodynamic equilibrium description, the solution to the appropriate Maxwell equations, and the solution of the linear transport equation describing neutral atom transport in plasmas. The specific problems of plasma heating by ion cyclotron radiofrequency (ICRF) waves and neutral atom transport in the plasma edge and in complicated geometrical components such as divertor channels or pumped limiter structures are analyzed. A theoretical, onedimensional slab model of ICRF heating at ω = 2ω/SUB cD/ is developed and applied to determine the space-time response of tokamak plasmas. Generally, strong single-pass absorption is found for high-density, high (β) plasmas using a low k 11 spectrum (0.05 to 0.1 cm -1 ) although for (β > 1%, electron Landau damping becomes important. Deterministic and Monte Carlo methods to solve the neutral atom transport problem are described. Specific application to determine the spectrum of neutral atoms emerging from the duct of a pump limiter shows it to be hard (mean energy > 20 eV), indicating very incomplete energy thermalization. Uncertainties are identified in the overall problem of dynamic burning plasma analysis caused by the complexity of the problem itself and by uncertainties in fundamental areas such as plasma transport coefficients, stability, and plasma edge physics
Transport, chaos and plasma physics
International Nuclear Information System (INIS)
Benkadda, S.; Doveil, F.; Elskens, Y.
1993-01-01
This workshop made it possible to gather for the first time plasma physicists, dynamical systems physicists and mathematicians, around a general theme focusing on the characterisation of chaotic transport. The participations have been divided into 4 topics: - dynamical systems and microscopic models of chaotic transport, - magnetic fluctuations and transport in tokamaks, - drift wave turbulence, self-organisation and intermittency, and - Wave-particle interactions
Hamiltonian analysis of fast wave current drive in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Becoulet, A; Fraboulet, D; Giruzzi, G; Moreau, D; Saoutic, B [Association Euratom-CEA, Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Chinardet, J [CISI Ingenierie, Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France)
1993-12-01
The Hamiltonian formalism is used to analyze the direct resonant interaction between the fast magnetosonic wave and the electrons in a tokamak plasma. The intrinsic stochasticity of the electron phase space trajectories is derived, and together with extrinsic de-correlation processes, assesses the validity of the quasilinear approximation for the kinetic studies of fast wave current drive (FWCD). A full-wave resolution of the Maxwell-Vlasov set of equations provides the exact pattern of the wave fields in a complete tokamak geometry, for a realistic antenna spectrum. The local quasilinear diffusion tensor is derived from the wave fields, and is used for a computation of the driven current and deposited power profiles, the current drive efficiency, including possible non-linear effects in the kinetic equation. Several applications of FWCD on existing and future machines are given, as well as results concerning combination of FWCD with other non inductive current drive methods. An analytical expression for the current drive efficiency is given in the high single-pass absorption regimes. (authors). 20 figs., 1 tab., 26 refs.
Hamiltonian analysis of fast wave current drive in tokamak plasmas
International Nuclear Information System (INIS)
Becoulet, A.; Fraboulet, D.; Giruzzi, G.; Moreau, D.; Saoutic, B.
1993-12-01
The Hamiltonian formalism is used to analyze the direct resonant interaction between the fast magnetosonic wave and the electrons in a tokamak plasma. The intrinsic stochasticity of the electron phase space trajectories is derived, and together with extrinsic de-correlation processes, assesses the validity of the quasilinear approximation for the kinetic studies of fast wave current drive (FWCD). A full-wave resolution of the Maxwell-Vlasov set of equations provides the exact pattern of the wave fields in a complete tokamak geometry, for a realistic antenna spectrum. The local quasilinear diffusion tensor is derived from the wave fields, and is used for a computation of the driven current and deposited power profiles, the current drive efficiency, including possible non-linear effects in the kinetic equation. Several applications of FWCD on existing and future machines are given, as well as results concerning combination of FWCD with other non inductive current drive methods. An analytical expression for the current drive efficiency is given in the high single-pass absorption regimes. (authors). 20 figs., 1 tab., 26 refs
Modelisation of synchrotron radiation losses in realistic tokamak plasmas
International Nuclear Information System (INIS)
Albajar, F.; Johner, J.; Granata, G.
2000-08-01
Synchrotron radiation losses become significant in the power balance of high-temperature plasmas envisaged for next step tokamaks. Due to the complexity of the exact calculation, these losses are usually roughly estimated with expressions derived from a plasma description using simplifying assumptions on the geometry, radiation absorption, and density and temperature profiles. In the present article, the complete formulation of the transport of synchrotron radiation is performed for realistic conditions of toroidal plasma geometry with elongated cross-section, using an exact method for the calculation of the absorption coefficient, and for arbitrary shapes of density and temperature profiles. The effects of toroidicity and temperature profile on synchrotron radiation losses are analyzed in detail. In particular, when the electron temperature profile is almost flat in the plasma center, as for example in ITB confinement regimes, synchrotron losses are found to be much stronger than in the case where the profile is represented by its best generalized parabolic approximation, though both cases give approximately the same thermal energy contents. Such an effect is not included in present approximate expressions. Finally, we propose a seven-variable fit for the fast calculation of synchrotron radiation losses. This fit is derived from a large database, which has been generated using a code implementing the complete formulation and optimized for massively parallel computing. (author)
Development of the 'JFT-2' tokamak plasma position control system
International Nuclear Information System (INIS)
Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.
1980-01-01
Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)
Spectra of heliumlike krypton from tokamak fusion test reactor plasmas
International Nuclear Information System (INIS)
Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M.; Smith, A.; Fraenkel, B.
1993-04-01
Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 Angstrom including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport
Tokamak-7 operation in experiments with a plasma
International Nuclear Information System (INIS)
Buzanki, V.V.; Bychkov, A.V.; Denisov, V.F.
1982-01-01
The results of experiments with plasma at the Tokamak-7 (T-7) device are presented. The experiments have been carried out with a constant diaphragm of 31,5 cm radius and two movable graphite diaphragms at the 26-28 cm plasma filament radius and 1,6-1,9 T magnetic field. Two stable regimes with 150 and 200 kA and 250 ms discharge current length have been investigated. It is shown that the strongest poloidal filed perturhations have been observed at the beginning of the discharge. Electron plasma temperature Tsub(e) has been determined from the spectrum analysis of soft X radiation by the foil method. Stable plasma regimes with current up to 200 kA, bypass voltage being equal 1,58V electron density -0,5-5,0 x 10 13 cm -3 , Tsub(e)=1,1-1,3 keV ion temperature-490 eV. The range between discharge pulses has reached 3 min. at the discharge current-240 kA. No considerable effect of magnetic field variables on the superconducting magnetic system has been observed
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)
Edge localized mode physics and operational aspects in tokamaks
International Nuclear Information System (INIS)
Becoulet, M; Huysmans, G; Sarazin, Y; Garbet, X; Ghendrih, Ph; Rimini, F; Joffrin, E; Litaudon, X; Monier-Garbet, P; Ane, J-M; Thomas, P; Grosman, A; Parail, V; Wilson, H; Lomas, P; Vries, P de; Zastrow, K-D; Matthews, G F; Lonnroth, J; Gerasimov, S; Sharapov, S; Gryaznevich, M; Counsell, G; Kirk, A; Valovic, M; Buttery, R; Loarte, A; Saibene, G; Sartori, R; Leonard, A; Snyder, P; Lao, L L; Gohil, P; Evans, T E; Moyer, R A; Kamada, Y; Chankin, A; Oyama, N; Hatae, T; Asakura, N; Tudisco, O; Giovannozzi, E; Crisanti, F; Perez, C P; Koslowski, H R; Eich, T; Sips, A; Horton, L; Hermann, A; Lang, P; Stober, J; Suttrop, W; Beyer, P; Saarelma, S
2003-01-01
Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios. Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs. High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high beta p are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and beta p . The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U, DIII-D) in
Edge localized mode physics and operational aspects in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Becoulet, M [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Huysmans, G [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Sarazin, Y [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Garbet, X [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Ghendrih, Ph [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Rimini, F [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Joffrin, E [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Litaudon, X [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Monier-Garbet, P [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Ane, J-M [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Thomas, P [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Grosman, A [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Parail, V [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Wilson, H [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Lomas, P [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Vries, P de[Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Zastrow, K-D [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Matthews, G F [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Lonnroth, J [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Gerasimov, S [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom)] [and others
2003-12-01
Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios. Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs. High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high beta{sub p} are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and beta{sub p}. The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U
The engineering design of the Tokamak Physics Experiment
International Nuclear Information System (INIS)
Schmidt, J.A.
1994-01-01
A mission and supporting physics objectives have been developed, which establishes an important role for the Tokamak Physics Experiment (TPX) in developing the physic basis for a future fusion reactor. The design of TPX include advanced physics features, such as shaping and profile control, along with the capability of operating for very long pulses. The development of the superconducting magnets, actively cooled internal hardware, and remote maintenance will be an important technology contribution to future fusion projects, such as ITER. The Conceptual Design and Management Systems for TPX have been developed and reviewed, and the project is beginning Preliminary Design. If adequately funded the construction project should be completed in the year 2000
Lyapunov-based distributed control of the safety-factor profile in a tokamak plasma
International Nuclear Information System (INIS)
Bribiesca Argomedo, Federico; Witrant, Emmanuel; Prieur, Christophe; Brémond, Sylvain; Nouailletas, Rémy; Artaud, Jean-François
2013-01-01
A real-time model-based controller is developed for the tracking of the distributed safety-factor profile in a tokamak plasma. Using relevant physical models and simplifying assumptions, theoretical stability and robustness guarantees were obtained using a Lyapunov function. This approach considers the couplings between the poloidal flux diffusion equation, the time-varying temperature profiles and an independent total plasma current control. The actuator chosen for the safety-factor profile tracking is the lower hybrid current drive, although the results presented can be easily extended to any non-inductive current source. The performance and robustness of the proposed control law is evaluated with a physics-oriented simulation code on Tore Supra experimental test cases. (paper)
International Nuclear Information System (INIS)
Viola, M.E.; McCann, J.
1985-01-01
During experimental operation, a problem developed with the mechanical integrity of the TFTR surface pumping system neutralizer plates that required a vacuum vessel entry for repairs. This problem, coupled with several less significant machine internal problems that had been developing, forced the decision to make an unscheduled vacuum vessel entry. An extended machine outage at that time would have had a severe impact on the experimental schedule. Therefore, the goal was to make repairs and return the vacuum vessel to a clean condition as quickly as possible. The total time required between the end of regularly scheduled activity and restoration of the machine capability to routinely obtain 1 MA disruption-free plasma was 12 days
Near-wall effects in improved plasma confinement regimes in tokamak FT-2
International Nuclear Information System (INIS)
Budnikov, V.N.; D'yachenko, V.V.; Esipov, L.A.
1997-01-01
Transition to the regime of improved plasma confinement (H-mode) revealed in experiments on low hybrid heating in tokamak ft-2 is analyzed. Main attention is paid to processes, taking place in near-wall region. The data are correlated with results of experiments in large tokamaks
Time-dependent analysis of the resistivity of post-disruption tokamak plasmas
International Nuclear Information System (INIS)
Bakhtiari, M.; Whyte, D. G.
2006-01-01
The effect of neutrals on plasma resistivity due to electron-neutral collisions is studied with respect to its effect on tokamak disruptions. The resistivity of the tokamak plasma after the thermal quench is critical in determining the current quench rate, the plasma temperature, and runaway electron generation in tokamaks through the electric field, all features which are important for mitigating the damaging effect of disruptions. It is shown that the plasma resistivity during tokamak disruptions is a time-dependent parameter which may vary with disruption time scales due to the increasing fraction of neutrals. However the effect of neutrals on resistivity is found to be small for the expected neutral fraction, mostly due to power balance considerations between radiation and Ohmic heating in the plasma
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
Plasma flow and transport on the tokamak ISTTOK boundary plasma
International Nuclear Information System (INIS)
Figueiredo, H.; Silva, C.; Goncalves, B.; Duarte, P.; Fernandes, H.
2011-01-01
The ISTTOK boundary plasma velocity near the outer midplane is measured on the parallel and perpendicular directions in four different configurations by reversing independently the toroidal magnetic field and the plasma current directions. The parallel flow is found to not depend significantly on both the toroidal magnetic field and plasma current directions, being always directed towards the nearest limiter in the scrape-off layer. On the contrary, the perpendicular flow is found to follow the E r x B drift direction. The poloidal velocity has also been derived from the correlation of floating potential signals measured on poloidally separated probes and a good agreement with the value derived with the Gundestrup probe is found. Finally, the dynamical interplay between parallel momentum and turbulent particle flux has been investigated and a clear dynamical coupling between these quantities is found in the region inside the limiter.
Determination of the plasma position for its real-time control in the COMPASS tokamak
International Nuclear Information System (INIS)
Janky, F.; Havlicek, J.; Valcarcel, D.; Hron, M.; Horacek, J.; Kudlacek, O.; Panek, R.; Carvalho, B.B.
2011-01-01
An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.
Determination of the plasma position for its real-time control in the COMPASS tokamak
Energy Technology Data Exchange (ETDEWEB)
Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Havlicek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Valcarcel, D. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal); Hron, M.; Horacek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Kudlacek, O. [Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, Technicka 2, 166 27 Prague (Czech Republic); Panek, R. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Carvalho, B.B. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal)
2011-10-15
An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.
Plasma density remote control system of experimental advanced superconductive tokamak
International Nuclear Information System (INIS)
Zhang Mingxin; Luo Jiarong; Li Guiming; Wang Hua; Zhao Dazheng; Xu Congdong
2007-01-01
In Tokamak experiments, experimental data and information on the density control are stored in the local computer system. Therefore, the researchers have to be in the control room for getting the data. Plasma Density Remote Control System (DRCS), which is implemented by encapsulating the business logic on the client in the B/S module, conducts the complicated science computation and realizes the synchronization with the experimental process on the client. At the same time, Web Services and Data File Services are deployed for the data exchange. It is proved in the experiments that DRCS not only meets the requirements for the remote control, but also shows an enhanced capability on the data transmission. (authors)
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
Sustained high βN plasmas on EAST tokamak
Gao, Xiang; the EAST team
2018-05-01
Sustained high normalized beta (βN ∼ 1.9) plasmas with an ITER-like tungsten divertor have been achieved on EAST tokamak recently. The high power NBI heating system of 4.8 MW and the 4.6 GHz lower hybrid wave of 1 MW were developed and applied to produce edge and internal transport barriers in high βN discharges. The central flat q profile with q (ρ) ∼ 1 at ρ safety factor q95 = 4.7 is identified by the multi-channel far-infrared laser polarimeter and the EFIT code. The fraction of non-inductive current is about 40%. The relation between fishbone activity and ITB formation is observed and discussed.
Modeling of noble gas injection into tokamak plasmas
International Nuclear Information System (INIS)
Morozov, D.Kh.; Yurchenko, E.I.; Lukash, V.E.; Baronova, E.O.; Rozhansky, V.A.; Senichenkov, I.Yu.; Veselova, I.Yu.; Schneider, R.
2005-01-01
Noble gas injection for mitigation of the disruption in DIII-D is simulated. The simulation of the first two stages is performed: of the neutral gas jet penetration through the background plasmas, and of the thermal quench. In order to simulate the first stage the 1.5-dimensional numerical code LLP with improved radiation model for noble gas is used. It is demonstrated that the jet remains mainly neutral and thus is able to penetrate to the central region of the tokamak in accordance with experimental observations. Plasma cooling at this stage is provided by the energy exchange with the jet. The radiation is relatively small, and the plasma thermal energy is spent mainly on the jet expansion. The magnetic surfaces in contact with the jet are cooled significantly. The cooling front propagates towards the plasma center. The simulations of the plasma column dynamics in the presence of moving jet is performed by means of the free boundary transport modeling DINA code. It has been shown that the cooling front is accompanied by strongly localized 'shark fin-like' perturbation in toroidal current density profile. After few milliseconds the jet (together with the current perturbation) achieves the region where safety factor is slightly higher than unity and a new type of the non-local kink mode develops. The unstable kink perturbation is non-resonant for any magnetic surface, both inside the plasma column, and in the vacuum space. The mode disturbs mainly the core region. The growth time of the 'shark fin-like' mode is higher than the Alfven time by a factor of 100 for DIII-D parameters. Hence, the simulation describes the DIII-D experimental results, at least, qualitatively. (author)
International Nuclear Information System (INIS)
Park, Hyeon, K.
1996-05-01
The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks
Plasma profile and shape optimization for the advanced tokamak power plant, ARIES-AT
International Nuclear Information System (INIS)
Kessel, C.E.; Mau, T.K.; Jardin, S.C.; Najmabadi, F.
2006-01-01
An advanced tokamak plasma configuration is developed based on equilibrium, ideal MHD stability, bootstrap current analysis, vertical stability and control, and poloidal field coil analysis. The plasma boundaries used in the analysis are forced to coincide with the 99% flux surface from the free-boundary equilibrium. Using an accurate bootstrap current model and external current drive profiles from ray tracing calculations in combination with optimized pressure profiles, β N values above 7.0 have been obtained. The minimum current drive requirement is found to lie at a lower β N of 6.0. The external kink mode is stabilized by a tungsten shell located at 0.33 times the minor radius and a feedback system. Plasma shape optimization has led to an elongation of 2.2 and triangularity of 0.9 at the separatrix. Vertical stability could be achieved by a combination of tungsten shells located at 0.33 times the minor radius and feedback control coils located behind the shield. The poloidal field coils were optimized in location and current, providing a maximum coil current of 8.6 MA. These developments have led to a simultaneous reduction in the power plant major radius and toroidal field from those found in a previous study [S.C. Jardin, C.E. Kessel, C.G. Bathke, D.A. Ehst, T.K. Mau, F. Najmabadi, T.W. Petrie, the ARIES Team, Physics basis for a reversed shear tokamak power plant, Fusion Eng. Design 38 (1997) 27
Thermal and nonthermal electron cyclotron emission by high-temperature tokamak plasmas
International Nuclear Information System (INIS)
Airoldi, A.; Ramponi, G.
1997-01-01
An analysis of the electron cyclotron emission (ECE) spectra emitted by a high-temperature tokamak plasma in the frequency range of the second and third harmonic of the electron cyclotron frequency is made, both in purely Maxwellian and in non-Maxwellian cases (i.e., in the presence of a current-carrying superthermal tail). The work is motivated mainly by the experimental observations made in the supershot plasmas of the Tokamak Fusion Test Reactor (TFTR), where a systematic disagreement is found between the T e measurements by second-harmonic ECE and Thomson scattering. We show that, by properly taking into account the overlap of superthermals-emitted third harmonic with second-harmonic bulk emission, the radiation temperature observed about the central frequency of the second harmonic may be enhanced up to 30%endash 40% compared to the corresponding thermal value. Moreover we show that, for parameters relevant to the International Thermonuclear Experimental Reactor (ITER) with T e (0)>7 keV, the overlap between the second and the downshifted third harmonic seriously affects the central plasma region, so that the X-mode emission at the second harmonic becomes unsuitable for local T e measurements. copyright 1997 American Institute of Physics
Numerical study of neoclassical plasma pedestal in a tokamak geometry
International Nuclear Information System (INIS)
Chang, C.S.; Ku, Seunghoe; Weitzner, H.
2004-01-01
The fundamental properties of steep neoclassical plasma pedestals in a quiescent tokamak plasma have been investigated with a new guiding center particle code XGC: an X-point included Guiding Center code. It is shown that the width of the steepest neoclassical pedestals is similar to an experimentally observed edge pedestal width, and that a steep pedestal must be accompanied by a self-consistent negative radial electric field well. It is also shown that a steep neoclassical pedestal can form naturally at a quiescent diverted edge as the particle source from the neutral penetration (and heat flux from the core plasma) is balanced by the sharply increasing convective ion loss toward the separatrix. The steep neoclassical pedestal and the strong radial electric field well are suppressed by an anomalous diffusion coefficient of a strength appropriate to an L-mode state; nonetheless, the ExB shearing rate increases rapidly with pedestal temperature. Additionally, the present study shows that a steep pedestal at the diverted edge acts as a cocurrent parallel momentum source
Plasma pressure in the discharge column of the Novillo Tokamak
International Nuclear Information System (INIS)
Gaytan G, E.
1995-01-01
The design and construction of an acquisition system for the measurement of the plasma pressure in the Novillo Tokamak is described in detail. The system includes a high voltage ramp generator, a hardware and a software interface with a personal computer. It is used to determine experimentally the variations of the pressure in the plasma column in the cleaning and main discharges. The measurement of the pressure is made with a Pirani sensor adapted to the acquisition hardware and synchronized with the discharge in the plasma. The software is made in object oriented programming as a graphic interface designed to be used easily. It controls the acquisition, records the data, displays in graphic form the results and save the measurements. The graphic interface is a building block that can be used in different acquisition tasks. The ramp generator can deliver a signal of 200 V peak to peak with a current of 200 m A and offset control. The acquisition time is 2.5 μ s for every measurement, 8192 measurements can be stored in the acquisition board for every discharge. (Author)
Analysis of IBW-driven plasma flows in tokamaks
International Nuclear Information System (INIS)
Berry, L.A.; Jaeger, E.F.; D'Azevedo, E.F.; Batchelor, D.B.; Carlsson, J.A.; Carter, M.D.; Cesario, R.
2001-01-01
Both theory and experiment have suggested that damping of Ion Bernstein Waves (IBWs) at ion cyclotron frequency harmonics could drive poloidal flows and lead to enhanced confinement for tokamaks. However, the early analyses were based on Reynolds stress closures of moment equations. More rigorous, finite Larmor radius (FLR) expansions of the radio frequency (RF) kinetic pressure for low harmonic interactions indicated that the Reynolds stress approximation was not generally valid, and resulted in significant changes in the plasma flow response. These changes were largest for wave interactions driven by finite Larmour radius effects. To provide a better assessment of higher harmonic interactions and IBW flow drive prospects, the electromagnetic (E and M) and RF kinetic force models are extended with no assumptions regarding the smallness of the ion Larmor radius. For both models, a spectral-width approximation was used to make the numerical analysis tractable. In addition, it was necessary to include the effects of plasma equilibrium gradients on the plasma conductivity and the RF-induced momentum in order to conserve energy and momentum. The analysis of high-harmonic IBW interactions for TFTR and FTU parameters indicates significant poloidal flow shears (relative to turbulence correlation times) for power levels available in present experiments. Recent advances in all-orders calculations of E and M fields in 2-D are also discussed. (author)
Plasma position control in a tokamak reactor around ignition
International Nuclear Information System (INIS)
Carretta, U.; Minardi, E.; Bacelli, N.
1986-01-01
Plasma position control in a tokamak reactor in the phase approaching ignition is closely related to burn control. If ignited burn corresponds to a thermally unstable situation the plasma becomes sensitive to the thermal instability already in the phase when ignition is approached so that the trajectory in the position-pressure (R,p) space becomes effectively unpredictable. For example, schemes involving closed cycles around ignition can be unstable in the heating-cooling phases, and the deviations may be cumulative in time. Reliable plasma control in pressure-position (p, R) space is achieved by beforehand constraining the p, R trajectory rigidly with suitable feedback vertical field stabilization, which is to be established already below ignition. A scheme in which ignition is approached in a stable and automatic way by feedback stabilization on the vertical field is proposed and studied in detail. The values of the gain coefficient ensuring stabilization and the associated p and R excursions are discussed both analytically, with a 0-D approximation including non-linear effects, and numerically with a 1-D code in cylindrical geometry. Profile effects increase the excursions, in particular above ignition. (author)
Perturbative transport experiments in JET Advanced Tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Mantica, P.; Gorini, G.; Sozzi, C. [Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Association, Milan (Italy); Imbeaux, F.; Sarazin, Y.; Garbet, X. [Association Euratom-CEA, St. Paul-lez-Durance Cedex (France); Kinsey, J. [Lehigh Univ., Bethlehem, Pennsylvania (United States); Budny, R. [Princeton Plasma Physics Lab, New Jersey (United States); Coffey, I.; Parail, V.; Walden, A. [Euratom/UKAEA Fusion Association, Abingdon, Oxon (United Kingdom); Dux, R. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Garzotti, L. [Istituto Gas Ionizzati, Padova (Italy); Ingesson, C. [FOM-Instituut voor Plasmafysica, Nieuwegein (Netherlands); Kissick, M. [University of California, Los Angeles (United States)
2003-07-01
Perturbative transport experiments have been performed in JET Advanced Tokamak plasmas either in conditions of fully developed Internal Transport Barrier (ITB) or during a phase where an ITB was not observed. Transient peripheral cooling was induced by either Laser Ablation or Shallow Pellet Injection and the ensuing travelling cold pulse was used to probe the plasma transport in the electron and, for the first time, also in the ion channel. Cold pulses travelling through ITBs are observed to erode the ITB outer part, but, if the inner ITB portion survives, it strongly damps the propagating wave. The result is discussed in the context of proposed possible pictures for ITB formation. In the absence of an ITB, the cold pulse shows a fast propagation in the outer plasma half, which is consistent with a region of stiff transport, while in the inner half it slows down but shows the peculiar feature of amplitude growing while propagating. The data are powerful tests for the validation of theoretical transport models. (author)
Runaway electrons in disruptions and perturbed magnetic topologies of tokamak plasmas
International Nuclear Information System (INIS)
Forster, Michael
2012-01-01
Nuclear fusion represents a valuable perspective for a safe and reliable energy supply from the middle of the 21st century on. Currently, the tokamak is the most advanced principle of confining a man-made fusion plasma. The operation of future, reactor sized tokamaks like ITER faces a crucial difficulty in the generation of runaway electrons. The runaway of electrons is a free fall acceleration into the relativistic regime which is known in various kinds of plasmas including astrophysical ones, thunderbolts and fusion plasmas. The tokamak disruption instability can include the conversion of a substantial part of the plasma current into a runaway electron current. When the high energetic runaways are lost, they can strike the plasma facing components at localised spots. Due to their high energies up to a few tens of MeV, the runaways carry the potential to reduce the lifetimes of wall components and even to destroy sensitive, i.e. actively cooled parts. The research for effective ways to suppress the generation of runaway electrons is hampered by the lack of a complete understanding of the physics of the runaways in disruptions. As it is practically impossible to use standard electron detectors in the challenging environment of a tokamak, the experimental knowledge about runaways is limited and it relies on rather indirect techniques of measurement. The main diagnostics used for this PhD work are three reciprocating probes which measure the runaway electrons directly at the plasma edge of the tokamak TEXTOR. A calorimetric probe and a material probe which exploits the signature that a runaway beam impact leaves in the probe were developed in the course of the PhD work. Novel observations of the burst-like runaway electron losses in tokamak disruptions are reported. The runaway bursts are temporally resolved and first-time measurements of the corresponding runaway energy spectra are presented. A characteristic shape and typical burst to burst variations of the
International conference on plasma physics
International Nuclear Information System (INIS)
Silin, V.P.; Sitenko, A.G.
1985-01-01
A brief report on the 6th International conference on plasma physics and on the 6th International Congress on plasma waves and plasma instabilities, which have taken place in summer 1984 in Losanne, is presented. Main items of the conference are enlightened, such as the general theory of a plasma, laboratory plasma, thermonuclear plasma, cosmic plasma and astrophysics
Selected highlights of ECH/ECCD physics studies in the TCV tokamak
Directory of Open Access Journals (Sweden)
Goodman T.P.
2015-01-01
Full Text Available The Tokamak a Configuration Variable, TCV, has used Electron Cyclotron Heating and Current Drive as its only auxiliary heating system for nearly two decades. In addition to basic plasma heating and current profiling, ECH and ECCD under either feedforward or real-time (feedback control allows control of plasma parameters and MHD behaviour to aid in physics studies and measurements. This paper describes four such studies in which EC control has proved crucial – increased resolution Thomson Scattering measurements in the plasma edge, time-resolved plasma rotation modification during the sawtooth cycle, robust neoclassical tearing mode (NTM suppression, and double pass transmission measurements of EC waves for scattering and polarization studies. The relative merits of feedforward and feedback methods for recent TCV experiments are discussed.
ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM
International Nuclear Information System (INIS)
HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG
2002-01-01
A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response
Experimental device for the X-ray energetic distribution measurement in a tokamak plasma
International Nuclear Information System (INIS)
Perez-Navarro, A.
1977-01-01
An experimental system to measure the X-ray spectrum in a tokamak plasma is described, emphasizing its characteristics: resolution, dead time and the pulse pile-up distortion effects on the X-ray spectra. (author) [es
Proceedings of the 21st symposium on plasma physics and technology
International Nuclear Information System (INIS)
Kulhanek, P.; Rezac, K.; Smetana, M.
2004-01-01
The supplement contains 159 papers out of the 229 papers presented at the conference; these papers were selected through the review process of the Czechoslovak Journal of Physics. The papers are divided into 5 categories corresponding to the main topics of the symposium, which covered all kinds of plasma research and associated applications: tokamaks and other magnetic confinement devices; short lived plasmas (plasma focus, z-pinch, X-ray sources); laser plasma; low temperature plasma; and plasma technology. All 22 papers dealing with tokamaks and other magnetic confinement devices were submitted to INIS as well as all 31 papers discussing short lived plasmas. (A.K.)
High density plasma heating in the Tokamak à configuration variable
International Nuclear Information System (INIS)
Curchod, L.
2011-04-01
The Tokamak à Configuration Variable (TCV) is a medium size magnetic confinement thermonuclear fusion experiment designed for the study of the plasma performances as a function of its shape. It is equipped with a high power and highly flexible electron cyclotron heating (ECH) and current drive (ECCD) system. Up to 3 MW of 2 nd harmonic EC power in ordinary (O 2 ) or extraordinary (X 2 ) polarization can be injected from TCV low-field side via six independently steerable launchers. In addition, up to 1.5 MW of 3 rd harmonic EC power (X 3 ) can be launched along the EC resonance from the top of TCV vacuum vessel. At high density, standard ECH and ECCD are prevented by the appearance of a cutoff layer screening the access to the EC resonance at the plasma center. As a consequence, less than 50% of TCV density operational domain is accessible to X 2 and X 3 ECH. The electron Bernstein waves (EBW) have been proposed to overcome this limitation. EBW is an electrostatic mode propagating beyond the plasma cutoff without upper density limit. Since it cannot propagate in vacuum, it has to be excited by mode conversion of EC waves in the plasma. Efficient electron Bernstein waves heating (EBH) and current drive (EBCD) were previously performed in several fusion devices, in particular in the W7-AS stellarator and in the MAST spherical tokamak. In TCV, the conditions for an efficient O-X-B mode conversion (i.e. a steep density gradient at the O 2 plasma cutoff) are met at the edge of high confinement (H-mode) plasmas characterized by the appearance of a pedestal in the electron temperature and density profiles. TCV experiments have demonstrated the first EBW coupling to overdense plasmas in a medium aspect-ratio tokamak via O-X-B mode conversion. This thesis work focuses on several aspects of ECH and EBH in low and high density plasmas. Firstly, the experimental optimum angles for the O-X-B mode conversion is successfully compared to the full-wave mode conversion calculation
International Nuclear Information System (INIS)
Pachtman, A.
1986-09-01
Measurements of Extraordinary mode absorption at the electron cyclotron harmonic frequencies are of unique value in high temperature, high density Tokamak plasma diagnostic applications. An experimental study of Extraordinary mode absorption at the semi-opaque second and third harmonics has been performed on the ALCATOR C Tokamak. A narrow beam of submillimeter laser radiation was used to illuminate the plasma in a horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission
Czech Academy of Sciences Publication Activity Database
Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Kraemer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Nanobashvili, S.; Peleman, P.; Razumova, K.A.; Stöckel, Jan; Vershkov, V.; Adámek, Jiří; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Brotánková, Jana; Dnestrovskij, A.Yu.; Ďuran, Ivan; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Hron, Martin; Ionita, C.; Jachmich, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Martines, E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Schrittwieser, R.; Soldatov, S.; Spolaore, M.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Žáček, František; Zimmermann, O.
2007-01-01
Roč. 49, 5A (2007), A29-A44 ISSN 0741-3335. [International Congress on Plasma Physics/13th./. Kiev, 22.05.2006-26.05.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * turbulence * electric fields Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 3.070, year: 2007
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
3D simulation studies of tokamak plasmas using MHD and extended-MHD models
International Nuclear Information System (INIS)
Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.
1996-01-01
The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data
Numerical study for determining PF coil system parameters in MHD equilibrium of KT-2 tokamak plasma
International Nuclear Information System (INIS)
Ryu, J.; Hong, S.H.; Lee, K.W.; Hong, B.G.; In, S.R.; Kim, S.K.
1995-01-01
The KT-2 is a large-aspect-ratio medium-sized divertor tokamak in the conceptual design phase and planned to be operational in 1998 at the Korea Atomic Energy Research Institute (KAERI). Plasma equilibrium in tokamak can be acquired by controlling the current of poloidal field (PF) coils in appropriate geometries and positions. In this study, the authors have performed numerical calculations to achieve the various equilibrium conditions fitting given plasma shapes and satisfying PF current limitations. Usually an ideal magnetohydrodynamic (MHD) equation is used to obtain the equilibrium solution of tokamak plasma, and it is practical to take advantage of a numerical method in solving the MHD equation because it has nonlinear source terms. Two equilibrium codes have been applied to find a double-null configuration of free-boundary tokamak plasma in KT-2: one is of the authors' own developing and the other is a free-boundary tokamak equilibrium code (FBT) that has been used mainly for the verification of developed code's results. PF coil system parameters including their positions and currents are determined for the optimization of input power required when the specifications of KT-2 tokamak are met. Then, several sets of equilibrium conditions during the tokamak operation are found to observe the changes of poloidal field currents with the passing of operation time step, and the basic stability problems related with the magnetic field structure is also considered
Anomalous electron streaming due to electrostatic modes in tokamak plasmas
International Nuclear Information System (INIS)
Schultz, S.D.; Bers, A.; Ram, A.K.
1993-01-01
The motion of circulating electrons in a tokamak interacting with electrostatic waves (such as lower-hybrid waves) is given by a guiding center Hamiltonian and studied by numerical integration. The unperturbed motion of electron guiding centers is first shown to be integrable, and, in a manner similar to that used in previous works, a set of action-angle coordinates for the orbits are derived which take into account finite aspect ratio and noncircular plasma cross section. Electrostatic modes in the low-frequency, long-wavelength limit are treated as a perturbation to the guiding center Hamiltonian. The waves are generated with low integral values of the toroidal and poloidal mode numbers n and m and satisfy the approximate lower-hybrid dispersion relation k perpendicular /k parallel ∼ ω pe /ω ∼ 10 1.5 . If the number of modes is greater than three, the electron motion parallel to the magnetic field is observed to be stochastic in the phase-space region where v parallel is near the wave parallel phase velocity. On surfaces with rational values of the safety factor q, superposition of modes with degenerate values of the parallel mode number n + (m/q) is shown to result in electron streaming perpendicular to the magnetic field. The speed and direction of this radial motion are observed to have sinusoidal dependence on the poloidal angle. For models including finite magnetic-field shear, the authors find a limit to the extent of the radial streaming of the electrons. Results for the speed of the electron radial motion for typical tokamak parameters are presented
Kinetic modelling of runaway electron avalanches in tokamak plasmas
International Nuclear Information System (INIS)
Nilsson, E; Peysson, Y; Saint-Laurent, F; Decker, J; Granetz, R S; Vlainic, M
2015-01-01
Runaway electrons can be generated in tokamak plasmas if the accelerating force from the toroidal electric field exceeds the collisional drag force owing to Coulomb collisions with the background plasma. In ITER, disruptions are expected to generate runaway electrons mainly through knock-on collisions (Hender et al 2007 Nucl. Fusion 47 S128–202), where enough momentum can be transferred from existing runaways to slow electrons to transport the latter beyond a critical momentum, setting off an avalanche of runaway electrons. Since knock-on runaways are usually scattered off with a significant perpendicular component of the momentum with respect to the local magnetic field direction, these particles are highly magnetized. Consequently, the momentum dynamics require a full 3D kinetic description, since these electrons are highly sensitive to the magnetic non-uniformity of a toroidal configuration. For this purpose, a bounce-averaged knock-on source term is derived. The generation of runaway electrons from the combined effect of Dreicer mechanism and knock-on collision process is studied with the code LUKE, a solver of the 3D linearized bounce-averaged relativistic electron Fokker–Planck equation (Decker and Peysson 2004 DKE: a fast numerical solver for the 3D drift kinetic equation Report EUR-CEA-FC-1736, Euratom-CEA), through the calculation of the response of the electron distribution function to a constant parallel electric field. The model, which has been successfully benchmarked against the standard Dreicer runaway theory now describes the runaway generation by knock-on collisions as proposed by Rosenbluth (Rosenbluth and Putvinski 1997 Nucl. Fusion 37 1355–62). This paper shows that the avalanche effect can be important even in non-disruptive scenarios. Runaway formation through knock-on collisions is found to be strongly reduced when taking place off the magnetic axis, since trapped electrons can not contribute to the runaway electron population. Finally
Resistive effects on helicity-wave current drive generated by Alfven waves in tokamak plasmas
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.; Komoshvili, K.
1997-01-01
This work is concerned with the investigation of non-ideal (resistive) MHD effects on the excitation of Alfven waves by externally launched fast-mode waves, in simulated tokamak plasmas; both continuum range, CR ({ω Alf (r)} min Alf (r)} max ) and discrete range, DR, where global Alfven eigenmodes, GAEs (ω Alf (r)} min ) exist, are considered. (Here, ω Alf (r) ≡ ω Alf [n(r), B 0 (r)] is an eigenfrequency of the shear Alfven wave). For this, a cylindrical current carrying plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell is used. Toroidicity effects are simulated by adopting for the axial equilibrium magnetic field component a suitable radial profile; shear and finite relative poloidal magnetic field are properly accounted for. A dielectric tensor appropriate to the physical conditions considered in this paper is derived and presented. (author)
Nonlinear transport processes in tokamak plasmas. I. The collisional regimes
International Nuclear Information System (INIS)
Sonnino, Giorgio; Peeters, Philippe
2008-01-01
An application of the thermodynamic field theory (TFT) to transport processes in L-mode tokamak plasmas is presented. The nonlinear corrections to the linear ('Onsager') transport coefficients in the collisional regimes are derived. A quite encouraging result is the appearance of an asymmetry between the Pfirsch-Schlueter (P-S) ion and electron transport coefficients: the latter presents a nonlinear correction, which is absent for the ions, and makes the radial electron coefficients much larger than the former. Explicit calculations and comparisons between the neoclassical results and the TFT predictions for Joint European Torus (JET) plasmas are also reported. It is found that the nonlinear electron P-S transport coefficients exceed the values provided by neoclassical theory by a factor that may be of the order 10 2 . The nonlinear classical coefficients exceed the neoclassical ones by a factor that may be of order 2. For JET, the discrepancy between experimental and theoretical results for the electron losses is therefore significantly reduced by a factor 10 2 when the nonlinear contributions are duly taken into account but, there is still a factor of 10 2 to be explained. This is most likely due to turbulence. The expressions of the ion transport coefficients, determined by the neoclassical theory in these two regimes, remain unaltered. The low-collisional regimes, i.e., the plateau and the banana regimes, are analyzed in the second part of this work
Modeling of the equilibrium of a tokamak plasma
International Nuclear Information System (INIS)
Grandgirard, V.
1999-12-01
The simulation and the control of a plasma discharge in a tokamak require an efficient and accurate solving of the equilibrium because this equilibrium needs to be calculated again every microsecond to simulate discharges that can last up to 1000 seconds. The purpose of this thesis is to propose numerical methods in order to calculate these equilibrium with acceptable computer time and memory size. Chapter 1 deals with hydrodynamics equation and sets up the problem. Chapter 2 gives a method to take into account the boundary conditions. Chapter 3 is dedicated to the optimization of the inversion of the system matrix. This matrix being quasi-symmetric, the Woodbury method combined with Cholesky method has been used. This direct method has been compared with 2 iterative methods: GMRES (generalized minimal residual) and BCG (bi-conjugate gradient). The 2 last chapters study the control of the plasma equilibrium, this work is presented in the formalism of the optimized control of distributed systems and leads to non-linear equations of state and quadratic functionals that are solved numerically by a quadratic sequential method. This method is based on the replacement of the initial problem with a series of control problems involving linear equations of state. (A.C.)
Double internal transport barrier triggering mechanism in tokamak plasmas
International Nuclear Information System (INIS)
Dong, Jiaqi; Mou, Zongze; Long, Yongxing; Mahajan, Swadesh M.
2004-01-01
Sheared flow layers created by energy released in magnetic reconnection processes are studied with the magneto hydrodynamics (MHD), aimed at internal transport barrier (ITB) dynamics. The double tearing mode induced by electron viscosity is investigated and proposed as a triggering mechanism for double internal transport barrier (DITB) observed in tokamak plasmas with non-monotonic safety factor profiles. The quasi-linear development of the mode is simulated and the emphasis is placed on the structure of sheared poloidal flow layers formed in the vicinity of the magnetic islands. For viscosity double tearing modes, it is shown that the sheared flows induced by the mode may reach the level required by the condition for ITB formation. Especially, the flow layers are found to form just outside the magnetic islands. The scaling of the generated velocity with plasma parameters is given. Possible explanation for the experimental observations that the preferential formation of transport barriers in the proximity of low order rational surface is discussed. (author)
Plasma rotation and transport in MAST spherical tokamak
Field, A. R.; Michael, C.; Akers, R. J.; Candy, J.; Colyer, G.; Guttenfelder, W.; Ghim, Y.-c.; Roach, C. M.; Saarelma, S.; MAST Team
2011-06-01
The formation of internal transport barriers (ITBs) is investigated in MAST spherical tokamak plasmas. The relative importance of equilibrium flow shear and magnetic shear in their formation and evolution is investigated using data from high-resolution kinetic- and q-profile diagnostics. In L-mode plasmas, with co-current directed NBI heating, ITBs in the momentum and ion thermal channels form in the negative shear region just inside qmin. In the ITB region the anomalous ion thermal transport is suppressed, with ion thermal transport close to the neo-classical level, although the electron transport remains anomalous. Linear stability analysis with the gyro-kinetic code GS2 shows that all electrostatic micro-instabilities are stable in the negative magnetic shear region in the core, both with and without flow shear. Outside the ITB, in the region of positive magnetic shear and relatively weak flow shear, electrostatic micro-instabilities become unstable over a wide range of wave numbers. Flow shear reduces the linear growth rates of low-k modes but suppression of ITG modes is incomplete, which is consistent with the observed anomalous ion transport in this region; however, flow shear has little impact on growth rates of high-k, electron-scale modes. With counter-NBI ITBs of greater radial extent form outside qmin due to the broader profile of E × B flow shear produced by the greater prompt fast-ion loss torque.
Plasma control issues for an advanced steady state tokamak reactor
International Nuclear Information System (INIS)
Moreau, D.
2001-01-01
This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)
Plasma control issues for an advanced steady state tokamak reactor
International Nuclear Information System (INIS)
Moreau, D.; Voitsekhovitch, I.
1999-01-01
This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)
A consistency analysis on the tokamak reactor plasmas
International Nuclear Information System (INIS)
Fukuyama, A.; Itoh, S.-I.; Itoh, K.
1990-12-01
The parameter regime which simultaneously fulfills the various physics constraints are looked for in the case of ITER grade tokamaks. The consistency analysis code is applied. It is found that, if the energy confinement time reaches 1.6 times of the prediction of the L-mode scaling law, the Q-value of about 4 is possible for the full current drive operation at the input power P in of 100MW (Q is the ratio of fusion output and P in ). In the ignition mode, where half of the current is inductively sustained, Q approaches to 15 for this circulating power. If only the L-mode is realized, Q is about 1.5 for P in ≅100 MW. (author)
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)
Numerical simulation in plasma physics
International Nuclear Information System (INIS)
Samarskii, A.A.
1980-01-01
Plasma physics is not only a field for development of physical theories and mathematical models but also an object of application of the computational experiment comprising analytical and numerical methods adapted for computers. The author considers only MHD plasma physics problems. Examples treated are dissipative structures in plasma; MHD model of solar dynamo; supernova explosion simulation; and plasma compression by a liner. (Auth.)
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.
Microwave reflectrometry for electron density measurements in the TJ-1 tokamak plasma
International Nuclear Information System (INIS)
Anabitarte, E.; Bustamante, E.G.; Calderon, M.A.G.; Vegas, A.
1986-01-01
A study about microwave reflectometry to measure the outside profile of the electron plasma density on tokamak TJ-1 is presented. It is also presented the condition of applicability of this method after the characteristic parameters of the plasma and its resolution. The simulation of the plasma in laboratory by means of a metallic mirror causes the whole characterization of the reflectometer. (author)
IR and FIR laser polarimetry as a diagnostic tool in high-. beta. and Tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Pereira, D; Machida, M; Scalabrin, A
1986-03-01
The change of the polarization state of an electromagnetic wave (EMW) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the EMW allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-pinch plasmas (lambda approx. 10..mu..m). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-pinch/UNICAMP, using lasers developed at UNICAMP.
International Nuclear Information System (INIS)
Killeen, J.
1975-08-01
The behavior of a plasma confined by a magnetic field is simulated by a variety of numerical models. Some models used on a short time scale give detailed knowledge of the plasma on a microscopic scale, while other models used on much longer time scales compute macroscopic properties of the plasma dynamics. In the last two years there has been a substantial increase in the numerical modelling of fusion devices. The status of MHD, transport, equilibrium, stability, Vlasov, Fokker-Planck, and Hybrid codes is reviewed. These codes have already been essential in the design and understanding of low and high beta toroidal experiments and mirror systems. The design of the next generation of fusion experiments and fusion test reactors will require continual development of these numerical models in order to include the best available plasma physics description and also to increase the geometric complexity of the model. (auth)
Fundamentals of plasma physics
Bittencourt, J A
1986-01-01
A general introduction designed to present a comprehensive, logical and unified treatment of the fundamentals of plasma physics based on statistical kinetic theory. Its clarity and completeness make it suitable for self-learning and self-paced courses. Problems are included.
International Nuclear Information System (INIS)
Outten, C.A.
1992-01-01
Electron Cyclotron Resonance (ECR) plasmas have been employed to simulate the plasma conditions at the edge of a tokamak in order to investigate hydrogen/helium uptake in thin metal films. The process of microwave power absorption, important to characterizing the ECR plasma source, was investigated by measuring the electron density and temperature with a Langmuir probe and optical spectroscopy as a function of the magnetic field gradient and incident microwave power. A novel diagnostic, carbon resistance probe, provided a direct measure of the ion energy and fluence while measurements from a Langmuir probe were used for comparison. The Langmuir probe gave a plasma potential minus floating potential of 30 ± 5 eV, in good agreement with the carbon resistance probe result of ion energy ≤ 40 eV. The measured ion energy was consistent with the ion energy predicted from a model based upon divergent magnetic field extraction. Also, based upon physical sputtering of the carbon, the hydrogen fluence rate was determined to be 1 x 10 16 /cm 2 -sec for 50 Watts of incident microwave power. ECR hydrogen/helium plasmas were used to study preferential pumping of helium in candidate materials for tokamak pump-limiters: nickel, vanadium, aluminum, and nickel/aluminum multi-layers. Nickel and vanadium exhibited similar pumping capacities whereas aluminum showed a reduced capacity due to increased sputtering. A helium retention model based upon ion implantation ranges and sputtering rates agreed with the experimental data. A new multilayer/bilayer pumping concept showed improved pumping above that for single element films
International Nuclear Information System (INIS)
Hoshino, Katsumichi
1989-09-01
A study on the heating and diagnosis of tokamak plasma by electromagnetic waves of electron cyclotron range of frequency is summarized. The main results obtained are as follows. On the engineering and technology, the technology of injecting high frequency, large power millimeter waves into tokamak plasma was established by carrying out the design, manufacture and test of a 60 GHz, 400 kW high frequency heating system, and the design, manufacture and test of a heterodyne type electron cyclotron radiation multi-channel mealsuring system were carried out, and the technology of measuring the radiation from tokamak plasma with the time resolution of 10 μs in multi-channel was established. On nuclear fusion reactor core engineering and plasma physics, the high efficiency electron heating of tokamak plasma by the incidence of fundamental irregular and regular waves at electron cyclotron frequency was verified. The discovery and analysis of the heating by electrostatic waves arising due to mode transformation from electromagnetic waves in upper hybrid resonance layer were carried out. By the incidence of second harmonic waves, the high efficiency electron heating of tokamak plasma was verified, and the heating characteristics were clarified. And others. (K.I.) 179 refs
International Nuclear Information System (INIS)
Peters, M.
1996-01-01
In the first experiment the plasma current in the RTP tokamak is varied. Here the underlying idea was to check whether at a low plasma current, transport in the tokamak resembles transport in stellarators more than at higher currents. Secondly, experiments have been done to study the relation of the diffusivity χ to the temperature and its gradient in both W7-AS and RTP. In this case the underlying idea was to find the explanation for the phenomenon observed in both tokamaks and stellarators that the quality of the confinement degrades when more heating is applied. A possible explanation is that the diffusivity increases with the temperature or its gradient. Whereas in standard tokamak and stellarator experiments the temperature and its gradient are strongly correlated, a special capability of the plasma heating system of W7-AS and RTP can force them to decouple. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Peters, M
1996-01-16
In the first experiment the plasma current in the RTP tokamak is varied. Here the underlying idea was to check whether at a low plasma current, transport in the tokamak resembles transport in stellarators more than at higher currents. Secondly, experiments have been done to study the relation of the diffusivity {chi} to the temperature and its gradient in both W7-AS and RTP. In this case the underlying idea was to find the explanation for the phenomenon observed in both tokamaks and stellarators that the quality of the confinement degrades when more heating is applied. A possible explanation is that the diffusivity increases with the temperature or its gradient. Whereas in standard tokamak and stellarator experiments the temperature and its gradient are strongly correlated, a special capability of the plasma heating system of W7-AS and RTP can force them to decouple. (orig.).
Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak
Energy Technology Data Exchange (ETDEWEB)
Berta, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Anda, G.; Bencze, A.; Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Háček, P., E-mail: hacek@ipp.cas.cz [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Hron, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Kovácsik, A. [Wigner – RCP, HAS, Budapest (Hungary); Department of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Pánek, R. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Réfy, D.; Veres, G. [Wigner – RCP, HAS, Budapest (Hungary); Weinzettl, V. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)
2015-10-15
Highlights: • Li-BES detection system on the COMPASS tokamak is optimized observation system with high temporal resolution. • High sensitivity to low level light fluctuations. • Optics and detectors with electronics are placed in thermally stabilized compact box. • Fast deflection system allows us to measure background corrected electron density profiles on microsecond time-scale. - Abstract: A new Li beam emission spectroscopy (Li-BES) diagnostic system with a ∼ cm spatial resolution, and with beam energy ranging from 10 keV up to 120 keV and a 18 channel Avalanche photo diode (APD) detector system sampled at 2 MHz has been recently installed and tested on the COMPASS tokamak. This diagnostic allows to reconstruct density profile based on directly measured light profiles, and to follow turbulent behaviour of the edge plasma. The paper reports technical capabilities of this new system designed for fine spatio-temporal measurements of plasma electron density. Focusing on turbulence-induced fluctuation measurements, we demonstrate how physically relevant information can be extracted using the COMPASS Li-BES system.
Plasma density measurements on COMPASS-C tokamak from electron cyclotron emission cutoffs
International Nuclear Information System (INIS)
Chenna Reddy, D.; Edlington, T.
1996-01-01
Electron cyclotron emission (ECE) is a standard diagnostic in present day tokamak devices for temperature measurement. When the plasma density is high enough the emission at some frequencies is cut off. Of these cutoff frequencies, the first frequency to cut off depends on the shape of the density profile. If the density profile can be described by a few parameters, in some circumstances, this first cutoff frequency can be used to obtain two of these parameters. If more than two parameters are needed to describe the density profile, then additional independent measurements are required to find all the parameters. We describe a technique by which it is possible to obtain an analytical relation between the radius at which the first cutoff occurs and the profile parameters. Assuming that the shape of the profile does not change as the average density rises after the first cutoff, one can use the cutoffs at other frequencies to obtain the average density at the time of these cutoffs. The plasma densities obtained with this technique using the data from a 14 channel ECE diagnostic on COMPASS-C tokamak are in good agreement with those measured by a standard 2 mm interferometer. The density measurement using the ECE cutoffs is an independent measurement and requires only a frequency calibration of the ECE diagnostic. copyright 1996 American Institute of Physics
Plasma Physics Network Newsletter. No. 4
International Nuclear Information System (INIS)
1991-08-01
This, fourth, issue of the Newsletter contains a (i) contribution in the series of reports on national fusion programmes from Algeria; (ii) a letter from Dr J.A.M. de Villiers, manager: fusion studies, at the Atomic Energy Corporation of South Africa Limited, informing about the close-down of the small tokamak project there, and soliciting ways to use some manpower and supportive sources to salvage the wealth of information still left behind in the project, and offering, in the possible absence of such manpower and supportive sources, the entire facility for sale (specifications of the Tokoloshe Tokamak plus diagnostic systems are enclosed); (iii) the e-mail address of the Third World Plasma Research Network (TWPRN), namely: ''PLASNET.NERUS.PFC.MIT.EDU''; (iv) minutes of the TWPRN Steering Committee Meeting held in May 1991, at the I.C.T.P., Trieste, Italy; (v) a news item on the ITER Tokamak project; (vi) a reiteration of the announcement of the 14th IAEA International Conference on Plasma Physics and Controlled Nuclear Fusion Research, to be held in Wuerzburg, Germany, September 30 - October 7, 1992; (vii) a list of IAEA Technical Committee Meetings during 1991; (viii) the First Announcement of the V Latin American Workshop on Plasma Physics, to be held in Mexico City, July 21-30, 1992, accompanied with a call for papers; all correspondence on this conference should be addressed to: Dr. Julio Herrera, V LAWPP, ICN-UNAM, Apdo. Postal 70-543, Delegacion Coyoacan, 04510 Mexico, D.F. Mexico (e-mail: ''HERRE.UNAMVM1.BITNET''); (ix) the announcement for the Second South North International Workshop on Fusion Theory, Lisbon, Portugal, March 1993 (contact: Pr. Tito Mendonca, Centro de Electrodinamica, Instituto Superio Tecnico, 1096 Lisbon Codex, Portugal)
Fundamentals of Plasma Physics
International Nuclear Information System (INIS)
Cargill, P J
2007-01-01
The widespread importance of plasmas in many areas of contemporary physics makes good textbooks in the field that are both introductory and comprehensive invaluable. This new book by Paul Bellen from CalTech by and large meets these goals. It covers the traditional textbook topics such as particle orbits, the derivation of the MHD equations from Vlasov theory, cold and warm plasma waves, Landau damping, as well as in the later chapters less common subjects such as magnetic helicity, nonlinear processes and dusty plasmas. The book is clearly written, neatly presented, and each chapter has a number of exercises or problems at their end. The author has also thankfully steered clear of the pitfall of filling the book with his own research results. The preface notes that the book is designed to provide an introduction to plasma physics for final year undergraduate and post-graduate students. However, it is difficult to see many physics undergraduates now at UK universities getting to grips with much of the content since their mathematics is not of a high enough standard. Students in Applied Mathematics departments would certainly fare better. An additional problem for the beginner is that some of the chapters do not lead the reader gently into a subject, but begin with quite advanced concepts. Being a multi-disciplinary subject, beginners tend to find plasma physics quite hard enough even when done simply. For postgraduate students these criticisms fade away and this book provides an excellent introduction. More senior researchers should also enjoy the book, especially Chapters 11-17 where more advanced topics are discussed. I found myself continually comparing the book with my favourite text for many years, 'The Physics of Plasmas' by T J M Boyd and J J Sanderson, reissued by Cambridge University Press in 2003. Researchers would want both books on their shelves, both for the different ways basic plasma physics is covered, and the diversity of more advanced topics. For
High-Q plasmas in the TFTR tokamak
International Nuclear Information System (INIS)
Jassby, D.L.; Barnes, C.W.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Dylla, H.F.; Efthimion, P.C.; Fredrickson, E.D.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.C.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S.J.; LaMarche, P.H.; LeBlanc, B.; Mansfield, D.K.; Marmar, E.S.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.K.; Paul, S.F.; Pitcher, S.; Ramsey, A.T.; Redi, M.H.; Sabbagh, S.A.; Scott, S.D.; Snipes, J.; Stevens, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Terry, J.L.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.
1991-01-01
In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength S n and D--D fusion power gain Q DD are realized in the neutral-beam-fueled and heated ''supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, S n increases approximately as P 1.8 b . The highest-Q shots are characterized by high T e (up to 12 keV), T i (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad T e profiles, and lower Z eff . Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the ''carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, Q DD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness [n e (0)/left-angle n e right-angle] during the beam pulse. To date, the best fusion results are S n =5x10 16 n/sec, Q DD =1.85x10 -3 , and neutron yield=4.0x10 16 n/pulse, obtained at I p =1.6--1.9 MA and beam energy E b =95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of S n arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with P b
Optimization of tokamak plasma equilibrium shape using parallel genetic algorithms
International Nuclear Information System (INIS)
Zhulin An; Bin Wu; Lijian Qiu
2006-01-01
In the device of non-circular cross sectional tokamaks, the plasma equilibrium shape has a strong influence on the confinement and MHD stability. The plasma equilibrium shape is determined by the configuration of the poloidal field (PF) system. Usually there are many PF systems that could support the specified plasma equilibrium, the differences are the number of coils used, their positions, sizes and currents. It is necessary to find the optimal choice that meets the engineering constrains, which is often done by a constrained optimization. The Genetic Algorithms (GAs) based method has been used to solve the problem of the optimization, but the time complexity limits the algorithms to become widely used. Due to the large search space that the optimization has, it takes several hours to get a nice result. The inherent parallelism in GAs can be exploited to enhance their search efficiency. In this paper, we introduce a parallel genetic algorithms (PGAs) based approach which can reduce the computational time. The algorithm has a master-slave structure, the slave explore the search space separately and return the results to the master. A program is also developed, and it can be running on any computers which support massage passing interface. Both the algorithm and the program are detailed discussed in the paper. We also include an application that uses the program to determine the positions and currents of PF coils in EAST. The program reach the target value within half an hour and yield a speedup rate of 5.21 on 8 CPUs. (author)
Electromagnetic microinstabilities in tokamak plasmas using a global spectral approach
Energy Technology Data Exchange (ETDEWEB)
Falchetto, G. L
2002-03-01
Electromagnetic microinstabilities in tokamak plasmas are studied by means of a linear global eigenvalue numerical code. The code is the electromagnetic extension of an existing electrostatic global gyrokinetic spectral toroidal code, called GLOGYSTO. Ion dynamics is described by the gyrokinetic equation, so that ion finite Larmor radius effects are taken into account to all orders. Non adiabatic electrons are included in the model, with passing particles described by the drift-kinetic equation and trapped particles through the bounce averaged drift-kinetic equation. A low frequency electromagnetic perturbation is applied to a low -but finite- {beta}plasma (where the parameter {beta} identifies the ratio of plasma pressure to magnetic pressure); thus, the parallel perturbations of the magnetic field are neglected. The system is closed by the quasi-neutrality equation and the parallel component of Ampere's law. The formulation is applied to a large aspect ratio toroidal configuration, with circular shifted surfaces. Such a simple configuration enables one to derive analytically the gyrocenter trajectories. The system is solved in Fourier space, taking advantage of a decomposition adapted to the toroidal geometry. The major contributions of this thesis are as follows. The electromagnetic effects on toroidal Ion Temperature Gradient driven (ITG) modes are studied. The stabilization of these modes with increasing {beta}, as predicted in previous work, is confirmed. The inclusion of trapped electron dynamics enables the study of its coupling to the ITG modes and of Trapped Electron Modes (TEM) .The effects of finite {beta} are considered together with those of different magnetic shear profiles and of the Shafranov shift. The threshold for the destabilization of an electromagnetic mode is identified. Moreover, the global formulation yields for the first time the radial structure of this so-called Alfvenic Ion Temperature Gradient (AITG) mode. The stability of the
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
Energy Technology Data Exchange (ETDEWEB)
Kunz, W
1979-01-01
This study examines the poloidal magnetic field distribution of tokamak plasma, and the elucidation of the radial distribution of the toroidal plasma flow. A numerical and experimental determination of the poloidal field based on the Faraday effect is presented. A method is discussed for measuring the rotation of the polarization plane linear polarized electromagnetic radiation, by passing through a plasma magnetized in the direction of the radiation. The polarization behavior of a linear polarized wave passing through a tokamak plasma is presented theoretically for various wavelengths, along with the experimental investigation of a ferrite modulation procedure through the use of different far infrared detectors.
Total hydrogen and oxygen fluxes in the edge plasma of tokamaks
International Nuclear Information System (INIS)
Kastelewicz, H.
1988-01-01
A relativistic model of the edge plasma of tokamaks is described considering the primary neutral fluxes emitted from limiter and wall. The primary neutrals, which determine essentially the particle flux balance in the plasma edge, the scrape-off layer plasma and the particles adsorbed at limiter and wall are treated as separate subsystems which are iteratively coupled through the mutual particle sinks and sources. The model is used for the calculation of total hydrogen and oxygen fluxes in edge plasma of tokamaks. The results for different fractions of and contributions to the total fluxes are illustrated and discussed
Proceedings of the 1984 International Conference on plasma physics
International Nuclear Information System (INIS)
Tran, M.Q.; Verbeek, R.J.
1985-01-01
The 1984 ICPP, held in Lausanne, Switzerland, is the third biennial conference of the series ''International conferences on plasma physics''. A complete spectrum of current plasma physics from fusion devices to interstellar space was presented, even if most of the papers were of direct interest for fusion. The conference stressed the important role that ''basic plasma physics'' must play in fusion research. Recent theoretical and experimental developments in tokamaks, stellarators, mirrors, reversed field pinches, and other fusion devices were reported. The successful operation of two newly-built large tokamak devices, JET and TFTR, holds the promise that a host of new results of decisive importance for fusion research will become available in the next few years. This is the first part of the conference
Plasma physics and controlled nuclear fusion research 1990. V. 1
International Nuclear Information System (INIS)
1991-01-01
Volume 1 of the Proceedings of the Thirteenth International Conference on Plasma Physics and Controlled Nuclear Fusion Research contains papers given in two of the sessions: A and E. Session A contains the Artsimovich Memorial Lecture and papers on tokamaks; session E papers on plasma heating and current drive. The titles and authors of each paper are listed in the Contents. Abstracts accompany each paper. Refs, figs and tabs
Interaction of a spheromak-like compact toroid with a high beta spherical tokamak plasma
International Nuclear Information System (INIS)
Hwang, D.Q.; McLean, H.S.; Baker, K.L.; Evans, R.W.; Horton, R.D.; Terry, S.D.; Howard, S.; Schmidt, G.L.
2000-01-01
Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak. (author)
The physics of an ignited tokamak
International Nuclear Information System (INIS)
Troyon, F.
1991-01-01
A next step device which can demonstrate long burn with a high probability of success is possible, without speculating on the accumulation of favorable effects and ignoring the other ones. From the data base available today this appears possible only with large devices such ans NET, ITER or JIT. If such a device is to provide the needed database to go to DEMO it must be able to explore a sufficiently wide range of parameter. From our present knowledge this can be obtained by choosing a conservative value of the reference magnetic field away from technological limits and compensate with a larger size. There remains for all devices very serious physics problems to solve, the most acute being the exhaust which must be compatible with a good qualtity enhanced confinement regime and a sufficiently high wall load to make a reactor. (orig.)
Enhanced confinement phenomenology in magnetic fusion plasmas: Is it unique in physics?
International Nuclear Information System (INIS)
Dendy, R.O.
2002-01-01
There is substantial experimental evidence that simple diffusive models for turbulent transport are insufficient to produce all the confinement phenomena observed in tokamaks. This paper reports on the emerging linkage between rapid, nonlocal, nondiffusive transport and overall confinement phenomenology including edge pedestals, enhanced confinement, ELMs, and internal transport barriers. Modern statistical physics techniques are used to construct simple models that generate many of the distinctive elements of global tokamak confinement phenomenology. The similarities are deep and are quantified. These results imply that current observations of avalanching transport in tokamaks may be deeply linked to the fundamental global features of tokamak plasma confinement. (author)
Impurity screening in high density plasmas in tokamaks with a limiter configuration
International Nuclear Information System (INIS)
Ferro, C.; Zanino, R.
1992-01-01
Impurity screening in high density plasmas in tokamaks with a limiter configuration is investigated by means of a simple semi-analytical model. An iterative scheme is devised, in order to determine self-consistently the values of scrape-off layer thickness, edge electron density and temperature, and main plasma contamination parameter Z eff , as a function of given average electron density and temperature in the main plasma and given input power. The model is applied to the poloidal limiter case of the Frascati Tokamak Upgrade, and results are compared with experimental data. A reasonable agreement between the trends is found, emphasizing the importance of a high edge plasma density for obtaining a clean main plasma in limiter tokamaks. (orig.)
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
Princeton Plasma Physics Laboratory annual report, October 1, 1991--September 30, 1992
Energy Technology Data Exchange (ETDEWEB)
1992-12-31
This report discusses the following topics: Principal parameters achieved in experimental devices for fiscal year 1992; tokamak fusion test reactor; princeton beta experiment-modification; current drive experiment-upgrade; tokamak physics experiment/steady-state advanced tokamak; international thermonuclear experimental reactor; international collaboration; x-ray laser studies; plasma processing: Deposition and etching of thin films; pure electron plasma experiments; theoretical studies; tokamak modeling; high-field magnet project; engineering department; environment, safety, and health and quality assurance; technology transfer; office of human resources and administration; PPPL invention disclosures for fiscal year 1992; office of resource management; graduate education: plasma physics; graduate education: program in plasma science and technology; and science education program.
Princeton Plasma Physics Laboratory annual report, October 1, 1991--September 30, 1992
International Nuclear Information System (INIS)
1992-01-01
This report discusses the following topics: Principal parameters achieved in experimental devices for fiscal year 1992; tokamak fusion test reactor; princeton beta experiment-modification; current drive experiment-upgrade; tokamak physics experiment/steady-state advanced tokamak; international thermonuclear experimental reactor; international collaboration; x-ray laser studies; plasma processing: Deposition and etching of thin films; pure electron plasma experiments; theoretical studies; tokamak modeling; high-field magnet project; engineering department; environment, safety, and health and quality assurance; technology transfer; office of human resources and administration; PPPL invention disclosures for fiscal year 1992; office of resource management; graduate education: plasma physics; graduate education: program in plasma science and technology; and science education program
Preliminary Safety Analysis Report for the Tokamak Physics Experiment
International Nuclear Information System (INIS)
Motloch, C.G.; Bonney, R.F.; Levine, J.D.; Masson, L.S.; Commander, J.C.
1995-04-01
This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR
Physics constraints on tokamak edge operational space and extrapolation to ITER
International Nuclear Information System (INIS)
Igitkhanov, Yu.; Janeschitz, G.; Sugihara, M.; Pacher, H.D.; Post, D.E.; Pacher, G.W.; Pogutse, O.P.
1998-01-01
This paper emphasises the theoretical understanding of the physical processes in the edge tokamak plasma and their attendant uncertainties and constraints. The various operational boundaries are represented in the edge operational space (EOS) diagram, the space of edge density and temperature, defined at the top of the H-mode transport barrier. The EOS is governed by four boundaries representing physical constraints for the edge plasma parameters. The first boundary represents the onset of type I ELM instabilities in terms of a critical pressure gradient for MHD stability at the edge which defines the maximum pedestal temperature for a given density once the width of the H-mode transport barrier at the edge (pedestal width) is known. The ideal ballooning mode is a candidate for this instability. The second boundary defines the boundary between type III ELM's, which are probably resistive MHD modes, and the ELM-free region. (orig.)