Configuration studies for a small-aspect-ratio tokamak stellarator hybrid
International Nuclear Information System (INIS)
Carreras, B.A.; Lynch, V.E.; Ware, A.
1996-08-01
The use of modulated toroidal coils offers a new path to the tokamak-stellarator hybrids. Low-aspect-ratio configurations can be found with robust vacuum flux surfaces and rotational transform close to the transform of a reverse-shear tokamak. These configurations have clear advantages in minimizing disruptions and their effect and in reducing tokamak current drive needs. They also allow the study of low-aspect-ratio effects on stellarator confinement in small devices
Study of an optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak
Energy Technology Data Exchange (ETDEWEB)
Hong, Bong Guen, E-mail: bghong@jbnu.ac.kr [Department of Quantum System Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of); Kim, Hoseok [Department of Applied Plasma Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of)
2016-11-15
Highlights: • Optimum configuration of a transmutation reactor based on a low aspect ratio tokamak was found. • Inboard and outboard radial build are determined by plasma physics, engineering and neutronics constraints. • Radial build and equilibrium fuel cycle play a major role in determining the transmutation characteristics. - Abstract: We determine the optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak. For self-consistent determination of the radial build of the reactor components, we couple a tokamak systems analysis with a radiation transport calculation. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on neutron multiplication, the tritium-breeding ratio, and the power density. We show that the breeding blanket model has an effect on the radial build of a transmutation blanket. A burn cycle has to be determined to keep the fast neutron fluence plasma-facing material below its radiation damage limit. We show that the radial build of the transmutation reactor components and the equilibrium fuel cycle play a major role in determining the transmutation characteristics.
Large aspect ratio tokamak study
International Nuclear Information System (INIS)
Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Sardella, C.; Wiseman, G.W.
1979-01-01
The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured
Maximum entropy tokamak configurations
International Nuclear Information System (INIS)
Minardi, E.
1989-01-01
The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)
Large Aspect Ratio Tokamak Study
International Nuclear Information System (INIS)
Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.
1980-06-01
The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak
Initial DEMO tokamak design configuration studies
Energy Technology Data Exchange (ETDEWEB)
Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others
2015-10-15
Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.
Configuration and structural design of Compact Ignition Tokamak
International Nuclear Information System (INIS)
Brown, T.G.
1985-01-01
Viewgraphs are presented on the configuration and structural design of the Compact Ignition Tokamak, originally presented to the US/Japan Workshop on Next Step Machine Design. Items discussed in this presentation include: PPPL 0424 ref design; MIT LITE ref design; IGNITOR 1.01 M ref design; and IGNITOR 1.08 M press configuration
Stability of tokamak magnetic configuration with a poloidal divertor
International Nuclear Information System (INIS)
Bazaeva, A.V.; Bykov, V.E.; Georgievskii, A.V.; Kaminskii, A.O.; Peletminskaya, V.G.; Pyatov, V.H.
1979-02-01
This paper investigates instabilities in the preseparatrix region of a tokamak magnetic configuration with a poloidal divertor with respect to perturbations produced by various irregularities in the manufacturing of tokamak magnetic systems. A computer solution, a system of differential equations describing the behavior of a force line, showed that small perturbation amplitudes may be the cause of the stochastic instability of force lines in the preseparatrix region. This instability is responsible for a number of demands on the accuracy in the manufacturing of tokamak magnetic systems. In particular, the misalignment in the divertor ring must not be larger than 0.5 0 , its displacement must be less than Δ/R = 10 -2 (Δ/R -2 ). This study can be used in the design of large thermonuclear installations
The energy principle applied to diverted tokamak configurations
International Nuclear Information System (INIS)
Atanasiu, C. V.; Guenter, S.; Lackner, K.; Moraru, A.; Zakharov, L. E.; Subbotin, A. A.
2008-01-01
Writing the expression of the potential energy in terms of the perturbation of the flux function, and performing an Euler minimisation, one obtains a system of ordinary differential equations in that perturbation. For a diverted configuration, the usual vanishing boundary conditions for the perturbed flux function at the magnetic axis and at infinity can no longer be used. In place of the vanishing boundary conditions at infinity, an approach to fix 'natural' boundary conditions for the system of differential equations for the perturbed flux function, just at the plasma boundary has been developed. As an example of application of our approaches, a particular equilibrium configuration of the ASDEX Upgrade tokamak has been considered and a detailed investigation of the dependence of the tearing stability parameter Δ' on plasma shape is given for a realistic tokamak equilibrium. The results shown are at least in qualitative agreement with experimental observations on ASDEX Upgrade and JET of a stabilizing influence of triangularity. The knowledge of Δ' for realistic tokamak plasmas is especially important for understanding of the plasma stability against NTMs. (authors)
Tokamak configuration analysis with the method of toroidal multipoles
International Nuclear Information System (INIS)
Micozzi, P.; Alladio, F.; Crisanti, F.; Marinucci, M.; Tanga, A.
1989-01-01
In the study of tokamak machines able to sustain plasmas of thermonuclear interest (JIT, IGNITOR, NET, CIT, ET), there is a strong quest for engineering optimization of the circuital components close to the plasma. We have developed a semianalytical axisymmetric MHD equilibrium code based on the technique of the poloidal ψ flux function expansion in toroidal harmonic series. This code is able to optimize the necessary currents in the poloidal circuits in order to sustain a plasma of fixed shape (also x-point configuration), toroidal current and poloidal β. (author) 4 refs., 4 figs
Second regime tokamak operation at large aspect ratio
International Nuclear Information System (INIS)
Navratil, G.A.
1989-01-01
This paper reviews the need for high beta in economic tokamak reactors and summarizes recent results on the scaling of the second regime beta limit for high-n ballooning modes using optimized pressure profiles as well as results on low-n mode stability at the first regime beta limit from the Columbia HBT tokamak. While several experiments have studied ballooning limits using high εβ p plasmas, the most important question for the use of the second stability regime for tokamak reactor improvement is how to achieve these high values of εβ p while at the same time increasing the value of beta to several times the Troyon beta limit. An approach to the study of this key question on beta limits using modest sized, large aspect ratio tokamaks is described. (author). 28 refs, 7 figs, 1 tab
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
Stability of high β large aspect ratio tokamaks
International Nuclear Information System (INIS)
Cowley, S.C.
1991-10-01
High β(β much-gt ε/q 2 ) large aspect ratio (ε much-gt 1) tokamak equilibria are shown to be always stable to ideal M.H.D. modes that are localized about a flux surface. Both the ballooning and interchange modes are shown to be stable. This work uses the analytic high β large aspect ratio tokamak equilibria developed by Cowley et.al., which are valid for arbitrary pressure and safety factor profiles. The stability results make no assumption about these profiles or the shape of the boundary. 14 refs., 4 figs
Omniclassical Diffusion in Low Aspect Ratio Tokamaks
International Nuclear Information System (INIS)
Mynick, H.E.; White, R.B.; Gates, D.A.
2004-01-01
Recently reported numerical results for axisymmetric devices with low aspect ratio A found radial transport enhanced over the expected neoclassical value by a factor of 2 to 3. In this paper, we provide an explanation for this enhancement. Transport theory in toroidal devices usually assumes large A, and that the ratio B p /B t of the poloidal to the toroidal magnetic field is small. These assumptions result in transport which, in the low collision limit, is dominated by banana orbits, giving the largest collisionless excursion of a particle from an initial flux surface. However in a small aspect ratio device one may have B p /B t ∼ 1, and the gyroradius may be larger than the banana excursion. Here, we develop an approximate analytic transport theory valid for devices with arbitrary A. For low A, we find that the enhanced transport, referred to as omniclassical, is a combination of neoclassical and properly generalized classical effects, which become dominant in the low-A, B p /B t ∼ 1 regime. Good agreement of the analytic theory with numerical simulations is obtained
Mechanical configuration for a superconducting ignition tokamak (TIBER)
International Nuclear Information System (INIS)
Neef, W.S. Jr.; Johnston, B.M.
1985-10-01
The Lawrence Livermore National Laboratory is evaluating the engineering feasibility and economics of a superconducting ignition tokamak. Two major operational requirements had to be satisfied: (1) the conductive heat leak to the refrigerated structure had to be minimized, and (2) assembly and maintenance of the entire experiment had to be possible with remotely operated tools. The middle poloidal ''push coil'' must have many annular disks to transfer the TF-coil inward force to the post without crushing superconductor. The toppling moment on the TF-coil vertical legs is huge. A method of keying together the TF-coil cases has been developed. This forms an integrated structure that resists torque. The joining technique permits linear motion for simple assembly/disassembly. the topping moment on the outer vertical legs of the TF coils is very large. To react that moment and avoid great coil-case bulk, we have developed a method that allows the PF coil support structure to assist the TF case structure. Finite element techniques were used to determine the ability of the coal case and conductor to react the magnetic loads. The entire cold-coil structure is mounted on a circular plate that is suspended by several large tension rods, similar to the MFTF-B yin-yang support rods. The vacuum vessel is all at room temperature and is configured like a bell jar with sixteen side doors, one for each shield module
Helical-tokamak hybridization concepts for compact configuration exploration and MHD stabilization
International Nuclear Information System (INIS)
Oishi, T.; Yamazaki, K.; Arimoto, H.; Baba, K.; Hasegawa, M.; Ozeki, H.; Shoji, T.; Mikhailov, M.I.
2010-11-01
To search for low-aspect-ratio torus systems, a lot of exotic confinement concepts are proposed so far historically. One of the authors previously proposed the tokamak-helical hybrid called TOKASTAR (Tokamak-Stellarator Hybrid) to improve the magnetic local shear near the bad curvature region. This is characterized by simple and compact coil systems with enough divertor space relevant to reactor designs. Based on this TOKASTAR concept, a toroidal mode number N=2 C (compact) -TOKASTAR machine (R - 35 mm) was constructed. The rotational transform of this compact helical configuration is rather small to confine hot ions, but can be utilized as a compact electron plasma machine for multi-purposes. The C-TOKASTAR has a pair of spherically winding helical coils and a pair of poloidal coils. Existence of magnetic surface and electron confinement property in C-TOKASTAR device were investigated by an electron-emission impedance method. Calculation of the particle orbit also supports that closed magnetic surface is formed in the cases that the ratio between poloidal and helical coil current is appropriate. Another aspect of the research using TOKASTAR configuration includes the evaluation of the effect of the outboard helical field application to tokamak plasmas. It is considered that outboard helical field has roles to assist the initiation of plasma current, to improve MHD stability, and so on. To check these roles, we made TOKASTAR-2 machine (R - 0.12 m, B - 1 kG) with ohmic heating central coil, eight toroidal field coils, a pair of vertical field coils and two outboard helical field coil segments. The electron cyclotron heating plasma start-up and plasma current disruption control experiments might be expected in this machine. Calculation of magnetic field line tracing has revealed that magnetic surface can be formed using additional outer helical coils. (author)
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
International Nuclear Information System (INIS)
Takizuka, Tomonori
2003-01-01
The formation mechanism of the current hole in tokamak plasmas is reviewed. Experimental results of JT-60U are shown. Increase of the off-central noninductive current is a key factor for the current-hole formation. The internal Transport Barrier (ITB), which generates large bootstrap current, plays an important role. The central current density in the hole stays nearly 0. The idea of a new equilibrium for a tokamak plasma with a current hole is introduced. This equilibrium configuration called Axisymmetric Tri-Magnetic-Islands (ATMI) equilibrium', has three islands along the R direction (a central-negative-current island and side-positive-current islands). The equilibrium is stable with the elongation coils when the current in the ATMI region is limited to a small amount. The confinement properties of a current-hole configuration with box-type ITB is described. A scaling of the core poloidal beta inside the ITB, β p,core , is given as ε f β p,core approx. = 1, which suggests the equilibrium limit (ε f : inverse aspect ratio at the ITB foot). Though the core stored energy is little dependent on the heating power, the estimated heat diffusivity in the ITB region moderately correlates with a neoclassical diffusivity. (author)
Relation of vertical stability and aspect ratio in tokamaks
International Nuclear Information System (INIS)
Stambaugh, R.D.; Lao, L.L.; Lazarus, E.A.
1992-01-01
It is evaluated how the upper limit to plasma elongation κ, caused by vertical stability, varies with the aspect ratio A=R/a of the tokamak. Equilibria were generated with EFITD and the vertical stability was assessed by GATO. For a 'generic' tokamak with a superconducting wall conformal to the plasma shape and a distance 0.5 a away from the plasma edge and a constant current profile (q 0 =1.0, l i ≅1.0, q 95 =3.2) it is found that the maximum stable κ decreased only slowly from 2.65 at A=2.0 to 2.4 at A=6.0. To first order, a reasonable assumption in trade-off studies of new machine designs is no dependence of κ max on A. (author). Letter-to-the-editor. 13 refs, 3 figs, 1 tab
Development path of low aspect ratio tokamak power plants
International Nuclear Information System (INIS)
Stambaugh, R.D.; Chan, V.S.; Miller, R.L.
1997-03-01
Recent advances in tokamak physics indicate the spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants. Full calculations of stability to kink and ballooning modes show the possibility of greater than 50% beta toroidal with the normalized beta as high as 10 and fully aligned 100% bootstrap current. Such beta values coupled with 2--3 T toroidal fields imply a pilot plant about the size of the present DIII-D tokamak could produce ∼ 800 MW thermal, 160 MW net electric, and would have a ratio of gross electric power over recirculating power (Q PLANT ) of 1.9. The high beta values in the ST mean that E x B shear stabilization of turbulence should be 10 times more effective in the ST than in present tokamaks, implying that the required high quality of confinement needed to support such high beta values will be obtained. The anticipated beta values are so high that the allowable neutron flux at the blanket sets the device size, not the physics constraints. The ST has a favorable size scaling so that at 2--3 times the pilot plant size the Q PLANT rises to 4--5, an economic range and 4 GW thermal power plants result. Current drive power requirements for 10% of the plasma current are consistent with the plant efficiencies quoted. The unshielded copper centerpost should have an adequate lifetime against nuclear transmutation induced resistance change and the low voltage, high current power supplies needed for the 12 turn TF coil appear reasonable. The favorable size scaling of the ST and the high beta mean that in large sizes, if the copper TF coil is replaced with a superconducting TF coil and a shield, the advanced fuel D-He 3 could be burned in a device with Q PLANT ∼ 4
Tight aspect ratio tokamak power reactor with superconducting TF coils
International Nuclear Information System (INIS)
Nishio, S.; Tobita, K.; Konishi, S.; Ando, T.; Hiroki, S.; Kuroda, T.; Yamauchi, M.; Azumi, M.; Nagata, M.
2003-01-01
Tight aspect ratio tokamak power reactor with super-conducting toroidal field (TF) coils has been proposed. A center solenoid coil system and an inboard blanket were discarded. The key point was how to find the engineering design solution of the TF coil system with the high field and high current density. The coil system with the center post radius of less than 1 m can generate the maximum field of ∼ 20 T. This coil system causes a compact reactor concept, where the plasma major and minor radii of 3.75 m and 1.9 m, respectively and the fusion power of 1.8 GW. (author)
Internal Magnetic Configuration Measured by ECE Imaging on EAST Tokamak
International Nuclear Information System (INIS)
Xu Ming; Wen Yizhi; Xie Jinlin; Yu Changxuan; Gao Bingxi; Xu Xiaoyuan; Liu Wandong; Hu Liqun; Sun Youwen; Qian Jinping; Wan Baonian
2013-01-01
ECE imaging (electron cyclotron emission imaging) is an important diagnostic which can give 2D imaging of temperature fluctuation in the core of tokamak. A method based on ECE imaging is introduced which can give the information of the position of magnetic axis and the structure of internal magnetic surface for EAST tokamak. The EFIT equilibrium reconstruction is not reliable due to the absence of important core diagnostic at the initial phase for EAST, so the information given by ECE imaging could help to improve the accuracy of EFIT equilibrium reconstruction. (magnetically confined plasma)
Survey of linear MHD stability in tokamak configurations
International Nuclear Information System (INIS)
Wakatani, M.
1977-01-01
The results found by MHD stability studies for both low-beta and high-beta tokamaks are reviewed. The stability against kink-ballooning modes in equilibria surrounded by vacuum or a layer of force free currents is considered. Internal kink modes and the relation to interchange modes, which should be considered after external kink modes are suppressed, are surveyed
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.)
Current drive and profile control in low aspect ratio tokamaks
International Nuclear Information System (INIS)
Chan, V.S.; Chiu, S.C.; Lin-Liu, Y.R.; Miller, R.L.; Turnbull, A.D.
1995-07-01
The key to the theoretically predicted high performance of a low aspect ratio tokamak (LAT) is its ability to operate at very large plasma current*I p . The plasma current at low aspect ratios follows the approximate formula: I p ∼ (5a 2 B t /Rqψ) [(1 + κ 2 )/2] [A/(A - 1)] where A quadruple-bond R/a which was derived from equilibrium studies. For constant qψ and B t , I p can increase by an order of magnitude over the case of tokamaks with A approx-gt 2.5. The large current results in a significantly enhanced β t (quadruple-bond β N I p /aB t ) possibly of order unity. It also compensates for the reduction in A to maintain the same confinement performance assuming the confinement time τ follows the generic form ∼ HI p P -1 / 2 R 3 / 2 κ 1 / 2 . The initiation and maintenance of such a large current is therefore a key issue for LATs
Engineering feasibility of tight aspect ratio Tokamak (spherical torus) reactors
International Nuclear Information System (INIS)
Peng, Y-K.M.; Hicks, J.B.
1990-01-01
Engineering solutions are identified and analyzed for key high-power-density components of tight aspect ratio tokamak reactors (spherical torus reactors). The potentially extreme divertor heat loads can be reduced to about 3 MW/m 2 in expanded divertors using coils inside the demountable toroidal field coils. Given the long and narrow divertor channels, gaseous divertor targets become possible, which eliminate sputtering and increase the divertor life. The unshielded centre conductor post (CCP) of the toroidal field coil can be made of a single dispersion strengthened copper conductor cooled by high-velocity pressurized water to maintain acceptable copper temperature and strength. Damage and activation of the CCP at a neutron fluence of 10 MW-a/m 2 are also tolerable. Annual replacement of the centre post, the divertor assemblies and the blanket can be accomplished with vertical access for all torus components, which are modularized to reduce size and weight. The technical requirements of these solutions are shown to be comparable with, if not less demanding than, those estimated for conventional tokamak reactors. (author)
Tokamak reactor designs as a function of aspect ratio
International Nuclear Information System (INIS)
Wong, C.P.C.; Stambaugh, R.D.
2000-01-01
This paper assesses the technical and economic potential of tokamak power plants which utilize superconducting coil (SC) or normal conducting coil (NC) designs as a function of aspect ratio (A). Based on the results from plasma equilibrium calculations, the key physics design parameters of β N , β p , β T , and κ were fitted to parametric equations covering A in the range of 1.2-6. By using ARIES-RS and ARIES-ST as reference design points, a fusion reactor system code was used to project the performance and cost of electricity (COE) of SC and NC reactor designs over the same range of A. The principle difference between the SC and the NC designs are the inboard standoff distance between the coil and the inboard first wall, and the maximum central column current density used for respective coil types. Results show that at an output power of 2 GWe both NC and SC designs can project COE in the respectable range of 62-65 mill/kW h at gross thermal efficiency of 46%, with neutron wall loading (Γ n ) ∼7 MW/m 2 . More importantly, we have learned that based on the present knowledge of equilibrium physics and fusion power core components and system design we can project the performance and COE of reactor designs at least for the purpose of comparative assessment. Tokamak design points can then be selected and optimized for testing or commercial devices as a function of output power, A and Γ n for both SC and NC design options
Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD
Energy Technology Data Exchange (ETDEWEB)
Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)
1994-11-01
A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.
Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD
International Nuclear Information System (INIS)
Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub; Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui; Chung, Kyoo Sun; Hong, Sang Heui; Kang, Heui Dong; Lee, Jae Koo
1994-11-01
A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the 'advanced tokamak' physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs
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.)
Linear tearing modes calculation for diverted tokamak configurations
International Nuclear Information System (INIS)
Atanasiu, C.V.; Guenter, S.; Lackner, K.; Moraru, A.; Zakharov, L.E.; Subbotin, A.A.
2004-01-01
Writing the expression of the potential energy in terms of the perturbation of the flux function, and performing an Euler minimization, one obtains a system of ordinary differential equations in that perturbation. For a diverted configuration, the usual vanishing boundary conditions for the perturbed flux function at the magnetic axis and at infinity can no longer be used. An approach to fix 'natural' boundary conditions for the perturbed flux function just at the plasma boundary has been developed; this replaces the vanishing boundary conditions at infinity. To obtain the necessary metric coefficients from an equilibrium solver, a function exhibiting the same singularity at the X point has been introduced, and thus the separatrix contour was described by a reduced number of moments
International Nuclear Information System (INIS)
Peng, Y.K.M.
1978-04-01
A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested
Configuration and engineering design of the ARIES-RS tokamak power plant
International Nuclear Information System (INIS)
Tillack, M.S.; Malang, S.; Waganer, L.; Wang, X.R.; Sze, D.K.; El-Guebaly, L.; Wong, C.P.C.; Crowell, J.A.; Mau, T.K.; Bromberg, L.
1997-01-01
ARIES-RS is a conceptual design study which has examined the potential of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. The design is a 1000 MWe, DT-burning fusion power plant based on the reversed-shear tokamak mode of plasma operation, and using moderately advanced engineering concepts such as lithium-cooled vanadium-alloy plasma-facing components. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. The ARIES-RS engineering design process emphasized the attainment of the top-level mission requirements developed in the early part of the study in a collaborative effort between the ARIES Team and representatives from U.S. electric utilities and industry. Major efforts were devoted to develop a credible configuration that allows rapid removal of full sectors followed by disassembly in the hot cells during plant operation. This was adopted as the only practical means to meet availability goals. Use of an electrically insulating coating for the self-cooled blanket and divertor provides a wide design window and simplified design. Optimization of the shield, which is one of the larger cost items, significantly reduced the power core cost by using ferritic steel where the power density and radiation levels are low. An additional saving is made by radial segmentation of the blanket, such that large segments can be reused. The overall tokamak configuration is described here, together with each of the major fusion power core components: the first-wall, blanket and shield; divertor; heating, current drive and fueling systems; and magnet systems. (orig.)
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)
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
A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium
International Nuclear Information System (INIS)
Reed, Mark; Parker, Ronald R.; Forget, Benoit
2012-01-01
This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more
Wave-driver options for low-aspect-ratio steady-state tokamak reactors
International Nuclear Information System (INIS)
Ehst, D.A.
1981-02-01
Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A = 2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value
A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks
International Nuclear Information System (INIS)
Woolley, R.D.
2009-01-01
A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress
A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks
International Nuclear Information System (INIS)
Woolley, Robert D.
2009-01-01
A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress.
Non-inductive current drive via helicity injection by Alfven waves in low aspects ratio Tokamak
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
1996-01-01
A theoretical investigation of radio frequency (RF) current drive via helicity injection in low aspect ratio tokamaks was carried out. A current-carrying cylindrical plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell was considered. Toroidal features of low aspect ratio tokamaks were simulated by incorporation of the following effects: (i) arbitrarily small aspect ratio, R o /a ≡ 1/ε (ii) strongly sheared equilibrium magnetic field; and (iii) relatively large poloidal component of the equilibrium magnetic field. The study concentrates on the Alfven continuum, i.e. the case in which the wave frequency satisfies the condition {ω Alf (r)} min ≤ω≥{ω Alf (r)} max , where ω Alf (r)≡ω[n(r),B o (o)] is an eigenfrequency of the shear Alfven wave (SAW). Thus, using low-p, ideal magneto-hydrodynamics, the wave equation with correct boundary (matching) conditions was solved, the RF field components were found and subsequently, current drive , power deposition and efficiency were computed. The results of our investigation clearly demonstrate the possibility of generation of RF-driven currents via helicity injection by Alfven waves in low aspect ratio tokamaks, in the SAW mode. A special algorithm was developed which enables the selection of the antenna parameters providing optimal current drive efficiency. (authors)
All Metal Iron Core For A Low Aspect Ratio Tokamak
International Nuclear Information System (INIS)
Gates, D.A.; Jun, C.; Zatz, I.; Zolfaghari, A.
2010-01-01
A novel concept for incorporating a iron core transformer within a axisymmetric toroidal plasma containment device with a high neutron flux is described. This design enables conceptual design of low aspect ratio devices which employ standard transformer-driven plasma startup by using all-metal high resistance separators between the toroidal field windings. This design avoids the inherent problems of a multiturn air core transformer which will inevitably suffer from strong neutron bombardment and hence lose the integrity of its insulation, both through long term material degradation and short term neutron-induced conductivity. A full 3-dimensional model of the concept has been developed within the MAXWELL program and the resultant loop voltage calculated. The utility of the result is found to be dependent on the resistivity of the high resistance separators. Useful loop voltage time histories have been obtained using achievable resistivities.
Characterization of peeling modes in a low aspect ratio tokamak
Bongard, M. W.; Thome, K. E.; Barr, J. L.; Burke, M. G.; Fonck, R. J.; Hinson, E. T.; Redd, A. J.; Schlossberg, D. J.
2014-11-01
Peeling modes are observed at the plasma edge in the Pegasus Toroidal Experiment under conditions of high edge current density (Jedge ˜ 0.1 MA m-2) and low magnetic field (B ˜ 0.1 T) present at near-unity aspect ratio. Their macroscopic properties are measured using external Mirnov coil arrays, Langmuir probes and high-speed visible imaging. The modest edge parameters and short pulse lengths of Pegasus discharges permit direct measurement of the internal magnetic field structure with an insertable array of Hall-effect sensors, providing the current profile and its temporal evolution. Peeling modes generate coherent, edge-localized electromagnetic activity with low toroidal mode numbers n ⩽ 3 and high poloidal mode numbers, in agreement with theoretical expectations of a low-n external kink structure. Coherent MHD fluctuation amplitudes are found to be strongly dependent on the experimentally measured Jedge/B peeling instability drive, consistent with theory. Peeling modes nonlinearly generate ELM-like, field-aligned filamentary structures that detach from the edge and propagate radially outward. The KFIT equilibrium code is extended with an Akima spline profile parameterization and an improved model for induced toroidal wall current estimation to obtain a reconstruction during peeling activity with its current profile constrained by internal Hall measurements. It is used to test the analytic peeling stability criterion and numerically evaluate ideal MHD stability. Both approaches predict instability, in agreement with experiment, with the latter identifying an unstable external kink.
Destruction of magnetic surfaces in the edge of a large aspect ratio Tokamak with ergodic limiter
International Nuclear Information System (INIS)
Viana, R.L.; Caldas, I.L.
1990-01-01
The model of Martin and Taylor for a large aspect-ratio Tokamak with an ergodic limiter is considered. In order to study the onset of chaotic behaviour for the magnetic field lines in the edge of the vessel, a Hamiltonian formulation is constructed for the system and the overlap of two peripheral magnetic islands is considered. So, it is possible to determine a threshold for the ergodic limiter current to cause destruction of rational magnetic surfaces in this region. (Author)
Resistive vs. total power depositions by Alfven modes in pre-heated low aspect ratio tokamaks
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
2004-01-01
The power deposition of fast waves launched by a LFS located antenna in a pre-heated, strongly non-uniform low aspect ratio tokamak (START) is investigated. The rigorous computational results indicate a total power deposition by far larger than that predicted for Alfven continuum eigenmodes in cylindrical plasmas. For toroidal wave numbers |N| > 1, the resistive and total power depositions are almost equal. (author)
International Nuclear Information System (INIS)
He Qibing; Peng Qiyang; Qu Wenxiao
1993-09-01
The effect of energetic trapped particles on the stabilization of ballooning modes in finite-aspect-ratio tokamaks is numerically analyzed. The numerical solution of boundary value problem of an integro-differential equation is successfully obtained by RKF integral method with variable step size. The results show that the instability domain of ballooning modes becomes small along with the increase of energetic particles pressure. The energetic trapped particles can partially or completely suppress the instability of ballooning modes
Profile formation and sustainment of autonomous tokamak plasma with current hole configuration
International Nuclear Information System (INIS)
Hayashi, N.; Takizuka, T.; Ozeki, T.
2005-01-01
We have investigated the profile formation and sustainment of tokamak plasmas with the current hole (CH) configuration by using 1.5D time-dependent transport simulations. A model of the current limit inside the CH on the basis of the Axisymmetric Tri-Magnetic-Islands equilibrium is introduced into the transport simulation. We found that a transport model with the sharp reduction of anomalous transport in the reversed-shear (RS) region can reproduce the time evolution of profiles observed in JT-60U experiments. The transport becomes neoclassical-level in the RS region, which results in the formation of profiles with internal transport barrier (ITB) and CH. The CH plasma has an autonomous property because of the strong interaction between a pressure profile and a current profile through the large bootstrap current fraction. The ITB width determined by the neoclassical-level transport agrees well with that measured in JT-60U. The energy confinement inside the ITB agrees with the scaling based on the JT-60U data. The scaling means the autonomous limitation of energy confinement in the CH plasma. The plasma with the large CH is sustained with the full current drive by the bootstrap current. The plasma with the small CH and the small bootstrap current fraction shrinks due to the penetration of inductive current. This shrink is prevented and the CH size can be controlled by the appropriate external current drive (CD). The CH plasma is found to respond autonomically to the external CD. (author)
Aspect Ratio Scaling of Ideal No-wall Stability Limits in High Bootstrap Fraction Tokamak Plasmas
International Nuclear Information System (INIS)
Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Maingi, R.; Sabbagh, S.A.; Soukhanovskii, V.; Stutman, D.
2003-01-01
Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40 (2000) 557] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profile-optimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime
International Nuclear Information System (INIS)
Bourbon, F.
1993-03-01
The present thesis treats of the modeling, of the numerical simulation and equilibrium optimal control and configuration evolution of the plasma in a tokamak. In the first chapter, we establish the problem of equilibrium of the plasma by the finite elements method and a Newton algorithm. Then, we investigate the problem of equilibrium control: determination of currents in circuits of the tokamak to reach a given configuration. In the third chapter, we develop a model for evolution of the plasma configuration during a discharge. We solve this question by a Newton algorithm; with the fourth chapter, we treat the monitoring of plasma configuration evolution
Finite Larmor radius effects on Alfven wave current drive in low-aspect ratio tokamaks
International Nuclear Information System (INIS)
Komoshvili, K.; Cuperman, S.; Bruma, C.
1998-01-01
Alfven wave current drive (AWCD) in low-aspect ratio (A≡R/a=1/ε > or approx. 1) tokamaks (LARTs) is studied numerically. For this, the full-wave equation (E parallel ≠0) with a Vlasov-based dielectric tensor is solved by relaxation techniques, subject to appropriate boundary conditions at the plasma centre and at the plasma-vacuum interface, as well as the concentric antenna current sheet and at the external metallic wall. A systematic investigation of the physical characteristics of the AWCD generated in LARTs when kinetic effects are considered is carried out and illustrative results are presented and discussed. (author)
International Nuclear Information System (INIS)
Hinnov, E.; Suckewer, S.; Cohen, S.; Sato, K.
1981-11-01
A number of spectrum lines of highly ionized copper, nickel, iron, chromium, and germanium have been observed and the corresponding transitions identified. The element under study is introduced into the discharge of the PLT Tokamak by means of rapid ablation by a laser pulse. The ionization state is generally distinguishable from the time behavior of the emitted light. New identifications of transitions are based on predicted wavelengths (from isoelectronic extrapolation and other data) and on approximate expected intensities. All the transitions pertain to the ground configurations of the respective ions, which are the only states strongly populated at tokamak plasma conditions. These lines are expected to be useful for spectroscopic plasma diagnostics in the 1-3 keV temperature range, and they provide direct measurement of intersystem energy separations from chromium through copper in the oxygen, nitrogen, and carbon isoelectronic sequences
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.
2003-01-01
We investigate quantitatively the possibility of turbulence suppression through the generation of transport barriers in pre-heated low aspect ratio tokamaks (LARTs) by the sheared electric fields generated by externally driven rf waves in the frequency range ω A ≅ ci (ω A and ω ci are, respectively, Alfven and ion cyclotron frequencies). To this aim the following sequential steps are followed: (1) Solutions of the resistive two-fluid model full wave equation for a realistic LART configuration (D-shape cross-section and aspect ratio, R/a ∼> 1; as well as suitably located low field side, LFS, antenna) upon using a quite general dielectric tensor operator; (2) Calculation of the ponderomotive forces and their magnetic surface averages; (3) Solution of a strongly non-linear differential equation for the quasi-stationary radial electric field, including the particle orbit squeezing effects, based on the results of steps (1) and (2); and (4) Calculation of the radial flow shear, S perpend ; for both banana and potato collisional regimes
Energy Technology Data Exchange (ETDEWEB)
Bruma, C.; Cuperman, S. [Tel Aviv Univ. (Israel). School of Physics and Astronomy; Komoshvili, K. [The College of Judea and Samaria, Ariel (Israel)
2003-05-01
We investigate quantitatively the possibility of turbulence suppression through the generation of transport barriers in pre-heated low aspect ratio tokamaks (LARTs) by the sheared electric fields generated by externally driven rf waves in the frequency range {omega}{sub A} {approx_equal} <{omega}{sub ci} ({omega}{sub A} and {omega}{sub ci} are, respectively, Alfven and ion cyclotron frequencies). To this aim the following sequential steps are followed: (1) Solutions of the resistive two-fluid model full wave equation for a realistic LART configuration (D-shape cross-section and aspect ratio, R/a {approx}> 1; as well as suitably located low field side, LFS, antenna) upon using a quite general dielectric tensor operator; (2) Calculation of the ponderomotive forces and their magnetic surface averages; (3) Solution of a strongly non-linear differential equation for the quasi-stationary radial electric field, including the particle orbit squeezing effects, based on the results of steps (1) and (2); and (4) Calculation of the radial flow shear, S{sub perpend}; for both banana and potato collisional regimes.
Directory of Open Access Journals (Sweden)
Izaskun Garrido
2016-08-01
Full Text Available Plasma stability is one of the obstacles in the path to the successful operation of fusion devices. Numerical control-oriented codes as it is the case of the widely accepted RZIp may be used within Tokamak simulations. The novelty of this article relies in the hierarchical development of a dynamic control loop. It is based on a current profile Model Predictive Control (MPC algorithm within a multiloop structure, where a MPC is developed at each step so as to improve the Proportional Integral Derivative (PID global scheme. The inner control loop is composed of a PID-based controller that acts over the Multiple Input Multiple Output (MIMO system resulting from the RZIp plasma model of the Tokamak à Configuration Variable (TCV. The coefficients of this PID controller are initially tuned using an eigenmode reduction over the passive structure model. The control action corresponding to the state of interest is then optimized in the outer MPC loop. For the sake of comparison, both the traditionally used PID global controller as well as the multiloop enhanced MPC are applied to the same TCV shot. The results show that the proposed control algorithm presents a superior performance over the conventional PID algorithm in terms of convergence. Furthermore, this enhanced MPC algorithm contributes to extend the discharge length and to overcome the limited power availability restrictions that hinder the performance of advanced tokamaks.
Comments on the asymptotic treatment of tokamak MHD-stability at large aspect ratio
International Nuclear Information System (INIS)
Rebhan, E.
1980-01-01
In the asymptotic treatment of tokamak MHD stability at small inverse aspect ratio epsilon, the special case of poloidal wave number m=0 has been treated improperly in the literature for both axisymmetric and non-axisymmetric modes. In axisymmetric stability, a contribution to the perturbational vacuum field is either omitted or cancelled. In a variational stability analysis this field contribution provides σ 2 W with a correction term proportional to (1nepsilon) -1 , which may change the asymptotic range of stability and improve agreement with numerical finite-aspect-ratio results. In non-axisymmetric stability, for the perturbational vacuum field of the m=0 modes, usually the wrong of two possible solutions is chosen. It is shown why in many cases this wrong choice has no consequences on the correctness of the stability results, and circumstances are pointed out under which consequences may arise. (author)
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.; Komoshvili, K.
1998-01-01
Several low aspect ratio (spherical) Tokamaks (ST's) are now in operation or under construction. These devices would permit cost-effective and attractive embodiment of future fusion reactors: they would provide high β, good confinement and steady state operation at modest field values. Now, a steady state reactor has to be sustained by non-inductively driven currents. Recently, the generation of non-inductive current drive by Alfven waves (AWCD) has been investigated theoretically within the framework of ideal (E p arallel=0) MHD and non-ideal, resistive (E p arallel≠0) MHD; however, in all these cases, the tokamak device consisted of a cylindrical plasma with simulated toroidal effects. Rather encouraging results have been obtained. In this work we further investigate AWCD in ST's as follows: (i) we use consistent equilibrium profiles with neoclassical conductivity corresponding to an ohmic START discharge; (ii) incorporate effects due to neoclassical conductivity in the elements of the resistive MHD dielectric tensor, in the solution of the full (E p arallel≠0) wave equation as well as in the calculation of AWCD; and (iii) carry out a systematic search for antenna parameters optimizing the AWCD. (author)
Energy Technology Data Exchange (ETDEWEB)
Bruma, C.; Cuperman, S.; Komoshvili, K. [School of Physics and Astronomy, Tel Aviv University, Tel Aviv (Israel)
1998-08-01
Several low aspect ratio (spherical) Tokamaks (ST's) are now in operation or under construction. These devices would permit cost-effective and attractive embodiment of future fusion reactors: they would provide high {beta}, good confinement and steady state operation at modest field values. Now, a steady state reactor has to be sustained by non-inductively driven currents. Recently, the generation of non-inductive current drive by Alfven waves (AWCD) has been investigated theoretically within the framework of ideal (E{sub p}arallel=0) MHD and non-ideal, resistive (E{sub p}arallel{ne}0) MHD; however, in all these cases, the tokamak device consisted of a cylindrical plasma with simulated toroidal effects. Rather encouraging results have been obtained. In this work we further investigate AWCD in ST's as follows: (i) we use consistent equilibrium profiles with neoclassical conductivity corresponding to an ohmic START discharge; (ii) incorporate effects due to neoclassical conductivity in the elements of the resistive MHD dielectric tensor, in the solution of the full (E{sub p}arallel{ne}0) wave equation as well as in the calculation of AWCD; and (iii) carry out a systematic search for antenna parameters optimizing the AWCD. (author)
International Nuclear Information System (INIS)
Stacey, W. M.
2001-01-01
Several series of model problem calculations have been performed to investigate the predicted effect of pumping, divertor configuration and fueling on the maximum achievable density in diverted tokamaks. Density limitations due to thermal instabilities (confinement degradation and multifaceted axisymmetric radiation from the edge) and to divertor choking are considered. For gas fueling the maximum achievable density is relatively insensitive to pumping (on or off), to the divertor configuration (open or closed), or to the location of the gas injection, although the gas fueling rate required to achieve this maximum achievable density is quite sensitive to these choices. Thermal instabilities are predicted to limit the density at lower values than divertor choking. Higher-density limits are predicted for pellet injection than for gas fueling
International Nuclear Information System (INIS)
Menon, M.M.; Mioduszewski, P.K.; Owen, L.W.; Anderson, P.M.; Baxi, C.B.; Langhorn, A.; Luxon, J.L.; Mahdavi, M.A.; Schaffer, M.J.; Schaubel, K.M.; "" class="author-name" title=" (General Atomics Co., San Diego, CA (United States))" data-affiliation=" (General Atomics Co., San Diego, CA (United States))" >Smith, J.P>
1992-01-01
In this paper, a particle exhaust scheme using a cryocondensation pump in the advanced divertor configuration of the DIII-D tokamak is described. In this configuration, the pump is located inside a baffle chamber within the tokamak, designed to receive particles reflected off the divertor strike region. A concentric coaxial loop with forced-convection flow of two-phase helium is selected as the cryocondensation surface. The pumping configuration is optimized by Monte Carlo techniques to provide maximum exhaust efficiency while minimizing the deleterious effects of impingement of energetic plasma particles on cryogenic surfaces. Heat loading contributions from various sources on the cryogenic surfaces are estimated, based on which the cryogenic surfaces are estimated, based on which the cryogenic flow loop for the pump is designed. The mechanical aspects of the pump, designed to meet the many challenging requirements of operating the cryopump internal to the tokamak vacuum and in close proximity with the high-temperature plasma, are also outlined
Stable equilibria for bootstrap-current-driven low aspect ratio tokamaks
International Nuclear Information System (INIS)
Miller, R.L.; Lin-Liu, Y.R.; Turnbull, A.D.; Chan, V.S.; Pearlstein, L.D.; Sauter, O.; Villard, L.
1997-01-01
Low aspect ratio tokamaks (LATs) can potentially provide a high ratio of plasma pressure to magnetic pressure β and high plasma current I at a modest size. This opens up the possibility of a high-power density compact fusion power plant. For the concept to be economically feasible, bootstrap current must be a major component of the plasma current, which requires operating at high β p . A high value of the Troyon factor β N and strong shaping is required to allow simultaneous operation at a high-β and high bootstrap fraction. Ideal magnetohydrodynamic stability of a range of equilibria at aspect ratio 1.4 is systematically explored by varying the pressure profile and shape. The pressure and current profiles are constrained in such a way as to assure complete bootstrap current alignment. Both β N and β are defined in terms of the vacuum toroidal field. Equilibria with β N ≥8 and β∼35%endash 55% exist that are stable to n=∞ ballooning modes. The highest β case is shown to be stable to n=0,1,2,3 kink modes with a conducting wall. copyright 1997 American Institute of Physics
Energy Technology Data Exchange (ETDEWEB)
Tal, B.; Nagy, D.; Veres, G. [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics, Hungarian Academy of Sciences, Association EURATOM, P. O. Box 49, H-1525 Budapest (Hungary); Labit, B.; Chavan, R.; Duval, B. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOM-Confédération Suisse, EPFL SB CRPP, Station 13, CH-1015 Lausanne (Switzerland)
2013-12-15
We introduce an upgraded version of a tomographical system which is built up from Absolute Extreme Ultraviolet-type (AXUV) detectors and has been installed on the Tokamak à Configuration Variable (TCV). The system is suitable for the investigation of fast radiative processes usually observed in magnetically confined high-temperature plasmas. The upgrade consists in the detector protection by movable shutters, some modifications to correct original design errors and the improvement in the data evaluation techniques. The short-term sensitivity degradation of the detectors, which is caused by the plasma radiation itself, has been monitored and found to be severe. The results provided by the system are consistent with the measurements obtained with the usual plasma radiation diagnostics installed on TCV. Additionally, the coupling between core plasma radiation and plasma-wall interaction is revealed. This was impossible with other available diagnostics on TCV.
International Nuclear Information System (INIS)
Lee, V.D.
1987-01-01
The Fusion Engineering Design Center (FEDC) is part of a national design team that is developing the conceptual design of the Compact Ignition Tokamak (CIT). To achieve a compact device with the minimum major radius, a vertical preload system is being developed to react the vertical separating force normally carried by the inboard leg of the toroidal field (TF) coils. The preload system is in the form of a hydraulic press. Challenges in the design include the development of hydraulic and structural systems for very large force requirements, which could interface with the CIT machine, while allowing maximum access to the top, bottom, and radial periphery of the machine. Maximum access is necessary for maintenance, diagnostics, instrumentation, and control systems. Materials used in the design must function in the nuclear environment and in the presence of high magnetic fields. This paper presents the configuration development of the hydraulic press used to vertically preload the CIT device
Basic toroidal Effects on Alfven Wave Current in Small Aspect Ratio Tokamaks
International Nuclear Information System (INIS)
Burma, C.; Cuperman, S.; Komoshvili, K.
1998-01-01
The Alfven wave current drive (AWCD) in small aspect ratio Tokamaks is properly calculated, with consideration of the basic toroidicity effects present in (i) the dielectric tensor-operator (involving the strongly toroidal equilibrium profiles), (ii) the structure of the r.f. fields obtained as a solution of the wave equation (through Maxwell's equations' toroidal operators as well as the conversion rate and conversion layer location, depending also on the equilibrium profiles) and (iii) the formulation of the AWCD (which, besides its dependence on the r.f. fields - affected by toroidicity as mentioned at points (i) and (ii) - also requires the equilibrium-magnetic-surface averaging of non-resonant forces involved). Thus, we consider consistent equilibrium profiles with neo-classical conductivity corresponding to an ohmic START-like discharge; use a resistive (anisotropic) MHD dielectric tensor-operator Edith practically no limitations, adequate to describe the plasma response in the pre-heated stage ; solve numerically the 2(1/2)D full- wave equation by the aid of an advanced finite element code developed in; and evaluate the AWCD by the aid of the recently proposed, quite general formulation holding in the case of strongly toroidal fusion devices and including contributions due to helicity injection, momentum transfer and plasma Bow. A general discussion of the results obtained in this work is presented
Non-inductive current drive via helicity injection by Alfven waves in low-aspect-ratio tokamaks
Energy Technology Data Exchange (ETDEWEB)
Cuperman, S.; Bruma, C.; Komoshvili, K. [Tel Aviv Univ. (Israel). Sackler Faculty of Exact Sciences
1996-08-01
A theoretical investigation of radio-frequency (RF) current drive via helicity injection in low aspect ratio tokamaks is carried out. A current-carrying cylindrical plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell is considered. Toroidal features of low-aspect-ratio tokamaks are simulated by incorporating the following effects: (i) arbitrarily small aspect ratio, R{sub O}/a ``identical to`` 1/{epsilon}; (ii) strongly sheared equilibrium magnetic field; and (iii) relatively large poloidal component of the equilibrium magnetic field. This study concentrates on the Alfven continuum, i.e. the case in which the wave frequency satisfies the condition {l_brace}{omega}{sub Alf}({tau}){r_brace}{sub min}{r_brace} {<=} {omega} {<=} {l_brace}{omega}{sub Alf}({tau}){r_brace}{sub max}, where {omega}{sub Alf}({tau}) ``identical to`` {omega}{sub Alf}[n({tau}), B{sub O}({tau})] is an eigenfrequency of the shear Alfven wave (SAW). Thus, using low-{beta} magnetohydrodynamics, the wave equation with correct boundary (matching) conditions is solved, the RF field components are found, and subsequently current drive, power deposition and efficiency are computed. The results of our investigation clearly demonstrate the possibility of generation of RF-driven currents via helicity injection by Alfven waves in low-aspect-ratio tokamaks, in the SAW mode. A special algorithm is developed that enables one to select the antenna parameters providing optimal current drive efficiency. (Author).
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.; Komoshvili, K.
1999-01-01
Some basic aspects of wave-plasma interaction of interest for tight aspect ratio spherical tokamaks are investigated theoretically. The following scenario is considered: A. Fast magnetosonic waves are launched by an external antenna into a simulated spherical Tokamak plasma; these waves are converted to Alfven waves at points (layer) satisfying the Alfven resonance condition. B. The simulated spherical tokamaks-plasma has a circular cross-section and toroidicity effects are simulated by Grad-Shafranov type, radially dependent axial magnetic field and its shear. (J. Actual equilibrium profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. D. The study is based on the numerical solution of the full e.m. wave equation which includes a quite general resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. Two kinds of results will be presented: I. Proofs validating the computational algorithm used and including convergence and energy conservation. II. Exact quantitative results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited p over . The dependence of the results on the launched wave characteristics (wave numbers, frequency and intensity) as well as on those of the equilibrium plasma (equilibrium current, neoclassical resistivity and electron inertia) will be discussed
Février, O.; Theiler, C.; De Oliveira, H.; Labit, B.; Fedorczak, N.; Baillod, A.
2018-05-01
This paper presents the current wall-embedded Langmuir probe system installed on the Tokamak à Configuration Variable (TCV), as well as the analysis tool chain used to interpret the current-voltage characteristic obtained when the probes are operated in swept-bias conditions. The analysis is based on a four-parameter fit combined with a minimum temperature approach. In order to reduce the effect of plasma fluctuations and measurement noise, several current-voltage characteristics are usually averaged before proceeding to the fitting. The impact of this procedure on the results is investigated, as well as the possible role of finite resistances in the circuitry, which could lead to an overestimation of the temperature. We study the application of the procedure in a specific regime, the plasma detachment, where results from other diagnostics indicate that the electron temperature derived from the Langmuir probes might be overestimated. To address this issue, we explore other fitting models and, in particular, an extension of the asymmetric double probe fit, which features effects of sheath expansion. We show that these models yield lower temperatures (up to approximately 60%) than the standard analysis in detached conditions, particularly for a temperature peak observed near the plasma strike point, but a discrepancy with other measurements remains. We explore a possible explanation for this observation, the presence of a fast electron population, and assess how robust the different methods are in such conditions.
International Nuclear Information System (INIS)
Lee, V.D.
1987-01-01
The Fusion Engineering Design Center (FEDC) is part of a national design team that is developing the conceptual design of the Compact Ignition Tokamak (CIT). To achieve a compact device with the minimum major radius, a vertical preload system is being developed to react the vertical separating force normally carried by the inboard leg of the toroidal field (TF) coils. The preload system is in the form of a hydraulic press. Challenges in the design include the development of hydraulic and structural systems for very large force requirements, which could interface with the CIT machine, while allowing maximum access to the top, bottom, and radial periphery of the machine. Maximum access is necessary for maintenance, diagnostics, instrumentation, and control systems. Materials used in the design must function in the nuclear environment and in the presence of high magnetic fields. The structural system developed is an arrangement in which the CIT device is installed in the jaws of the press. Large built-up beams above and below the CIT span the machine and deliver the vertical force to the center cylinder formed by the inboard legs of the TF coils. During the conceptual design study, the vertical force requirement has ranged between 25,000 and 52,000 t. The access requirement on top and bottom limits the width of the spanning beams. Nonmagnetic steel materials are also required because of operation in the high magnetic fields. In the hydraulic system design for the press, several options are being explored. These range from small-diameter jacks operating at very high pressure [228 MPa (33 ksi)] to large-diameter jacks operating at pressures up to 69 MPa (10 ksi). Configurations with various locations for the hydraulic cylinders have also been explored. The nuclear environment and maintenance requirements are factors that affect cylinder location. This paper presents the configuration development of the hydraulic press used to vertically preload the CIT device
Lee-side flow structures of very low aspect ratio cruciform wing–body configurations
CSIR Research Space (South Africa)
Tuling, S
2013-12-01
Full Text Available A numerical and experimental investigation was performed to study the dominant flow structures in the lee side of a cruciform wing–body configuration at supersonic speeds in the + orientation. The wings or strakes are of very low aspect ratio...
Stability of highly shifted equilibria in a large aspect ratio low-field tokamak
International Nuclear Information System (INIS)
Gourdain, P.-A.; Leboeuf, J.-N.; Neches, R. Y.
2007-01-01
In the long run, the economics of fusion will dictate that reactors confine large plasma pressure rather efficiently. A possible route manifests itself as equilibria with large shift of the plasma magnetic axis. This shift compresses the flux surfaces on the outer part of the plasma, hereby increasing the allowable plasma pressure a machine can confine for a given toroidal magnetic field, which is the main cost of the device. As a first step toward a reactor, we propose investigating the stability of such configurations in a low magnetic field high aspect ratio machine. By focusing our arguments solely on the shape of the toroidal plasma current density profile we discuss the stability of highly shifted equilibria and their robustness to current profile variations that could occur in actual experiments. The evolution of the plasma parameters, as the beta poloidal is increased, is also examined to give a better understanding of the difference in performance between the various regimes
International Nuclear Information System (INIS)
Leerink, S.; Heikkinen, J. A.; Janhunen, S. J.; Kiviniemi, T. P.; Nora, M.; Ogando, F.
2008-01-01
The ELMFIRE gyrokinetic simulation code has been used to perform full f simulations of the FT-2 tokamak. The dynamics of the radial electric field and the creation of poloidal velocity in the presence of turbulence are presented.
Evaluation of applicability of 2D iron core model for two-limb configuration of GOLEM tokamak
Czech Academy of Sciences Publication Activity Database
Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Vondrášek, G.
2013-01-01
Roč. 88, 6-8 (2013), s. 835-838 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Tokamak * Ferromagnetic core * Integral method * Tokamak GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613002573#
International Nuclear Information System (INIS)
Reid, R.L.; Barrett, R.J.; Brown, T.G.
1985-03-01
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged
International Nuclear Information System (INIS)
Robinson, D.C.
1987-01-01
With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.; Komoshvili, K.
1996-01-01
Supplementary non-inductive current drive and heating are necessary to bring Tokamak plasmas into the ignition regime. The resonant excitation of shear Alfven waves (SAW) - in the continuum range (CR) or/and in the discrete global Alfven eigenmode spectrum (GAE's) - represents one potential, suitable method for this purpose. Within the framework of ideal MHD, the current drive (CD) via helicity injection in Tokamak plasmas has been considered by Cuperman et al (1996) and Komoshvili et al. (1996). This work is concerned with the investigation of the non-ideal resistive MHD effects on both the excitation of SAW's (CR's and GAE's) and the generation of non-inductive current drive via helicity injection in Tokamak plasmas. The research covers Tokamak aspect ratios ranging between large value cases (R/a = 10) and the very tight value case (R/ a = 1.2). (authors)
Energy Technology Data Exchange (ETDEWEB)
Bruma, C.; Komoshvili, K. [Tel Aviv Univ. (Israel). School of Physics and Astronomy; Coll. of Judea and Samaria, Ariel (Israel); Cuperman, S. [Tel Aviv Univ. (Israel). School of Physics and Astronomy
2000-11-01
Some basic aspects of wave-plasma interaction of special interest for tight aspect ratio (spherical) tokamaks (ST's) are investigated numerically; these aspects include fast mode conversion and energy deposition. The study is based on the numerical solution of the full electro-magnetic (e.m.) wave equation which includes a quite general two-fluid, resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. A generalized expression for the power absorption appropriate for the above scenario, with consideration of all the basic effects also present in the dielectric tensor-operator, was derived and used. The current-carrying ST-plasma has a circular cross-section and toroidicity effects are simulated by a Grad-Shafranov type, radially dependent axial magnetic field and its shear; however, the Shafranov shift is not considered. Actually, the equilibrium parameters and radial profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. Fast magnetosonic waves are launched from an external antenna into this simulated spherical tokamak plasma; these waves are converted to Alfven waves at points (layers) satisfying the Alfven resonance condition. Quantitative-results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited power are presented. Their dependence on the equilibrium plasma current, neoclassical resistivity and electron inertia as well as on those of the antenna launched wave (wave numbers, frequency and current intensity) is systematically studied and discussed. (orig.)
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Kommoshvili, K [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Cuperman, S [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Bruma, C [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel)
2003-03-01
Kinetic effects in the conversion of fast waves to Alfven waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvenic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxiliary energy source for the successful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects.
International Nuclear Information System (INIS)
Bruma, C.; Komoshvili, K.; Cuperman, S.
2000-01-01
Some basic aspects of wave-plasma interaction of special interest for tight aspect ratio (spherical) tokamaks (ST's) are investigated numerically; these aspects include fast mode conversion and energy deposition. The study is based on the numerical solution of the full electro-magnetic (e.m.) wave equation which includes a quite general two-fluid, resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. A generalized expression for the power absorption appropriate for the above scenario, with consideration of all the basic effects also present in the dielectric tensor-operator, was derived and used. The current-carrying ST-plasma has a circular cross-section and toroidicity effects are simulated by a Grad-Shafranov type, radially dependent axial magnetic field and its shear; however, the Shafranov shift is not considered. Actually, the equilibrium parameters and radial profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. Fast magnetosonic waves are launched from an external antenna into this simulated spherical tokamak plasma; these waves are converted to Alfven waves at points (layers) satisfying the Alfven resonance condition. Quantitative-results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited power are presented. Their dependence on the equilibrium plasma current, neoclassical resistivity and electron inertia as well as on those of the antenna launched wave (wave numbers, frequency and current intensity) is systematically studied and discussed. (orig.)
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
International Nuclear Information System (INIS)
Kommoshvili, K; Cuperman, S; Bruma, C
2003-01-01
Kinetic effects in the conversion of fast waves to Alfven waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvenic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxiliary energy source for the successful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
Kommoshvili, K.; Cuperman, S.; Bruma, C.
2003-03-01
Kinetic effects in the conversion of fast waves to Alfvèn waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvènic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxilliary energy source for the succesful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects.
The basics of spherical tokamaks and progress in European research
International Nuclear Information System (INIS)
Gusev, V K; Alladio, F; Morris, A W
2003-01-01
When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper. European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference
Energy Technology Data Exchange (ETDEWEB)
Charles Kessel, et al
2014-03-05
The advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2 and triangularity of 0.63. The broadest pressure cases reached wall stabilized βN ~ 5.75, limited by n=3 external kink mode requiring a conducting shell at b/a = 0.3, and requiring plasma rotation, feedback, and or kinetic stabilization. The medium pressure peaking case reached βN = 5.28 with BT = 6.75, while the peaked pressure case reaches βN < 5.15. Fast particle MHD stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling show that about 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while over 95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring about ~ 1.1 MA of external current drive. This current is supplied with 5 MW of ICRF/FW and 40 MW of LHCD. EC was examined and is most effective for safety factor control over ρ ~ 0.2-0.6 with 20 MW. The pedestal density is ~ 0.9x1020 /m3 and the temperature is ~ 4.4 keV. The H98 factor is 1.65, n/nGr = 1.0, and the net power to LH threshold power is 2.8- 3.0 in the flattop.
Power plant design study of a high aspect ratio Tokamak using a SiC composite structure
International Nuclear Information System (INIS)
Murakami, Y.; Takase, H.; Shinya, K.
1998-01-01
The DREAM (drastically easy maintenance) tokamak is a fusion power plant which is designed from the viewpoint of maintenance feasibility. For this purpose, the DREAM reactor uses a plasma with a very high aspect ratio (A) and adopts SiC as a structural material. The choice of SiC affects the design of the core plasma, i.e. large inboard shield thickness, low synchrotron radiation reflectivity, and small plasma elongation for positional stability. The objectives of this study are to explore the feasibility of a high-A device, such as a power plant, and to clarify the technological impact of SiC material on the plasma design. Plasma size is optimized by the physics guidelines similar to ITER. The plasma major and minor radii of DREAM are 16 m and 2 m, respectively, and the average neutron wall load is 2.5 MW m -2 , the maximum toroidal field is 20 T, and the fusion power is 5.5 GW. Steady-state operation is obtained with 50 MW of external current-drive power and 90% bootstrap current. The divertor heat load is estimated to be about 10 MW m -2 . A radiative divertor concept is adopted to achieve a low divertor plasma temperature. The DREAM tokamak concept is found to be a possible candidate for a future power plant with more than 5 GW of fusion power and an acceptable divertor condition. (orig.)
Advanced commercial Tokamak optimization studies
International Nuclear Information System (INIS)
Whitley, R.H.; Berwald, D.H.; Gordon, J.D.
1985-01-01
Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units
Energy Technology Data Exchange (ETDEWEB)
Hong, Sang Hee; Lim, Ki Hang; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Duk Kyu [Seoul National University, Seoul (Korea, Republic of); Cho, Soo Won [Kyungki Unviersity, Suwon (Korea, Republic of)
1995-08-01
In the present project for developing the numerical codes of 2-DMHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we compute the plasma equilibrium of double null type and calculate the external coil currents and the plasma parameters used for operation and control data. Also the numerical algorithm is developed to analyse the behavior of edge plasmas in poloidal and radial directions and the programming and debugging of a 2-D transport code are completed. Furthermore, a neutral particle transport code for the edge region is developed and then used for the analysis of the neutral transport phenomena giving the sources in the fluid equations, and expected to supply the input parameters for the edge plasma transport code. 34 refs., 5 tabs., 28 figs. (author)
Technological and environmental prospects of low aspect ratio tokamak reactor VECTOR
International Nuclear Information System (INIS)
Nishio, S.; Tobita, K.; Isono, T.; Nakamura, Y.; Sato, M.; Sakurai, S.; Yamauchi, M.; Nishitani, T.; Tani, K.; Sengoku, S.; Kudo, Y.; Song, Y.; Tokimatsu, K.; Shinya, K.; Senda, I.; Konishi, S.
2005-01-01
Optimization of VECTOR design parameters has led to a fusion output of 2.5 GW with a small reactor weight of 8,800 tons. Current ramp-up simulation with TSC (Tokamak Simulation Code) demonstrated a stable current ramp-up with externally non-inductive current drive and bootstrap current when a central current is induced externally enough to avoid an extreme reversed shear such as current hole. It was numerically confirmed that such a low-A reactor would have an advantage of α-particle confinement. The divertor is designed to use a flux expansion of scrape-off-layer in the divertor region to maintain the heat flux on the divertor plate to be lower than 10 MW/m 2 without remote radiative cooling. Because of its compactness, CO 2 emission in a life cycle of a VECTOR power plant is estimated to be as low as 3.2 g-CO 2 /kWh, being lower than that of an ITER-sized DEMO reactor (4.9 g-CO 2 /kWh). As to the waste management of VECTOR, on the basis of reactor design and radiological considerations, we suggest reusing a liquid metal breeding material (PbLi) and neutron shield material (TiH 2 ) in successive reactors. Due to this waste management, its disposal waste would be reduced to as low as 3,000-4,000 tons, which is comparable with the radioactive waste of a light water reactor (4,000 tons in metal). (author)
International Nuclear Information System (INIS)
Burma, C.; Cuperman, S.; Komoshvili, K.
1998-01-01
The wave equation for strongly toroidal small aspect ratio (spherical) tokamaks with non-circular cross-section is properly formulated and solved for global waves, in the Alfven frequency range. The current-carrying toroidal plasma is surrounded by a helical sheet-current antenna, which is enclosed within a perfectly conducting wall. The problem is formulated in terms of the vector and scalar potentials (A,Φ), thus avoiding the numerical solution occurring in the case of (E,B) formulation. Adequate boundary conditions are applied at the vacuum - metallic wall interface and the magnetic axis. A recently derived dielectric tensor-operator, able to describe the anisotropic plasma response in spherical tokamaks, is used for this purpose; except for its linear character, no physical or geometrical limitations are imposed on it. The equilibrium profiles (magnetic field, pressure and current) are obtained from a numerical solution of the Grad-Shafranov equation. Specifically, the wave equation is solved by the aid of a numerical code we developed for the present problem, based on the well documented 2(1/2)D finite element solver proposed by E.G. Sewell. With the definitions V i (θ,ρ) = U i (-θ,ρ) (V i U i = A j , Φ; j = ρ,φ,θ), our code solves simultaneously 16 second order partial differential equations (eight equations for each of real and imaginary set of functions V i , U i ). A systematic analysis of the solutions obtained for various values and combinations of wavenumbers and frequencies in the Alfven range is presented
Energy Technology Data Exchange (ETDEWEB)
Viola, B.; Maddaluno, G.; Pericoli Ridolfini, V. [EURATOM-ENEA Association, C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Rome) (Italy); Corrigan, G.; Harting, D. [Culham Centre of Fusion Energy, EURATOM-Association, Abingdon (United Kingdom); Mattia, M. [Dipartimento di Informatica, Sistemi e Produzione, Universita di Roma, Tor Vergata, Via del Politecnico, 00133 Roma (Italy); Zagorski, R. [Institute of Plasma Physics and Laser Microfusion-EURATOM Association, 01-497 Warsaw (Poland)
2014-06-15
The new magnetic configurations for tokamak divertors, snowflake and super-X, proposed to mitigate the problem of the power exhaust in reactors have clearly evidenced the need for an accurate and reliable modeling of the physics governing the interaction with the plates. The initial effort undertaken jointly by ENEA and IPPLM has been focused to exploit a simple and versatile modeling tool, namely the 2D TECXY code, to obtain preliminary comparison between the conventional and snowflake configurations for the proposed new device FAST that should realize an edge plasma with properties quite close to those of a reactor. The very interesting features found for the snowflake, namely a power load mitigation much larger than expected directly from the change of the magnetic topology, has further pushed us to check these results with the more sophisticated computational tool EDGE2D coupled with the neutral code module EIRENE. After a preparatory work that has been carried out in order to adapt this code combination to deal with non-conventional, single null equilibria and in particular with second order nulls in the poloidal field generated in the snowflake configuration, in this paper we describe the first activity to compare these codes and discuss the first results obtained for FAST. The outcome of these EDGE2D runs is in qualitative agreement with those of TECXY, confirming the potential benefit obtainable from a snowflake configuration. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
Collisional Transport in a Low Aspect Ratio Tokamak -- Beyond the Drift Kinetic Formalism
International Nuclear Information System (INIS)
Gates, D.A.; White, R.B.
2004-01-01
Calculations of collisional thermal and particle diffusivities in toroidal magnetic plasma confinement devices order the toroidal gyroradius to be small relative to the poloidal gyroradius. This ordering is central to what is usually referred to as neoclassical transport theory. This ordering is incorrect at low aspect ratio, where it can often be the case that the toroidal gyroradius is larger than the poloidal gyroradius. We calculate the correction to the particle and thermal diffusivities at low aspect ratio by comparing the diffusivities as determined by a full orbit code (which we refer to as omni-classical diffusion) with those from a gyroaveraged orbit code (neoclassical diffusion). In typical low aspect ratio devices the omni-classical diffusion can be up to 2.5 times the calculated neoclassical value. We discuss the implications of this work on the analysis of collisional transport in low aspect ratio magnetic confinement experiments
Energy Technology Data Exchange (ETDEWEB)
Hong, Sang Hee; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Deok Kyu; Chung, TaeKyun; Chung, Mo Se [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)
1995-08-01
In the present project for developing the numerical codes of 2-D MHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we computed the MHD equilibria of single and double null configurations and determined the external coil currents and the plasma parameters used for operation and control data. Also we numerically acquired the distributions of edge plasma parameters in poloidal and radial directions= and the design-related values according to the various operating conditions using the developed plasma transport code. Furthermore, a neutral particle transport code for the edge region is developed and them used for the analysis of the neutral particle behavior yielding the source terms in the fluid transport equations, and expected to supply the input parameters for the edge plasma transport code. 53 refs., 12 tabs., 44 figs. (author)
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
1999-01-01
The authors developed a consistent formalism for the full wave equation, appropriate for the study of propagation, absorption and wave conversion of externally launched waves in strongly toroidal, spherical tokamaks with non-circular cross-section. This includes also the formulation of rigorous regularity, boundary, gauge and periodicity conditions suitable for the exact solution of the wave equation for such devices
International Nuclear Information System (INIS)
Furno, I.; Weisen, H.
2003-01-01
In the Tokamak a Configuration Variable [F. Hofmann, J.B. Lister, M. Anton et al., Plasma Phys. Controlled Fusion 36, B277 (1994)], inward or outward convection in the core of electron cyclotron heated and current driven plasmas is observed, depending on discharge conditions. In sawtoothing discharges with central electron cyclotron heating, outward convection is observed when a quasicontinuous m=1 kink mode is present, resulting in inverted sawteeth on the central electron density, while in the absence thereof, inward convection between successive sawtooth crashes leads to 'normal' sawteeth. The occurrence of a kink mode depends sensitively on plasma triangularity. When sawteeth are stabilized with central co- or counterelectron cyclotron current drive, stationary hollow electron density profiles are observed in the presence of m=1 modes, while peaked or flat profiles are observed in magnetohydrodynamic quiescent discharges. The observation of peaked density profiles in fully electron cyclotron driven plasmas demonstrates that pinch processes other than the Ware pinch must be responsible for these phenomena
Burke, M. G.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Hinson, E. T.; Perry, J. M.; Reusch, J. A.; Schlossberg, D. J.
2017-07-01
Plasmas in the Pegasus spherical tokamak are initiated and grown by the non-solenoidal local helicity injection (LHI) current drive technique. The LHI system consists of three adjacent electron current sources that inject multiple helical current filaments that can reconnect with each other. Anomalously high impurity ion temperatures are observed during LHI with T i,OV ⩽ 650 eV, which is in contrast to T i,OV ⩽ 70 eV from Ohmic heating alone. Spatial profiles of T i,OV indicate an edge localized heating source, with T i,OV ~ 650 eV near the outboard major radius of the injectors and dropping to ~150 eV near the plasma magnetic axis. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent injected current filaments. In these experiments, the HeII T i perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, while {{T}\\text{i,\\parallel}} experiences little change, in agreement with two-fluid reconnection theory. This ion heating is not expected to significantly impact the LHI plasma performance in Pegasus, as it does not contribute significantly to the electron heating. However, estimates of the power transfer to the bulk ion are quite large, and thus LHI current drive provides an auxiliary ion heating mechanism to the tokamak plasma.
Measurement of the fission ratio for several configurations of uranium oxide rod clusters
International Nuclear Information System (INIS)
Pattenden, S.K.; Patterson, C.R.
1962-02-01
This report describes measurements of the fission ratio for a single fuel channel of oxide rod clusters in an essentially infinite block of graphite. The measurements were made using the 'catcher-foil' technique, the activities of the catcher foils being measured by β-counting. Results are given, for 37-rod; 18-rod; 7-rod and 3-rod clusters, and are compared with theoretical predictions. (author)
HTMR: an experimental tokamak reactor with hybrid copper/superconductor toroidal field magnet
International Nuclear Information System (INIS)
Avanzini, P.G.; Raia, G.; Rosatelli, F.; Zampaglione, V.
1985-01-01
The feasibility of a hybrid configuration superconducting coils/copper coils for a next generation tokamak TF magnet has been investigated. On the basis of this hybrid solution, the conceptual design has been developed for a medium-high toroidal field tokamak reactor (HTMR). The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1
Directory of Open Access Journals (Sweden)
Adhi Tri Partono
2018-01-01
Full Text Available Ammonia is one of the most important industrial commodity due to its wide function. Ammonia synthesis reaction is an exotermic reaction. Therefore, Feed-Effluent Heat Exchanger (FEHE is added to increase thermal efficiency. However, FEHE could lead the process to experience hysteresis phenomenon due to interaction between equipments as one steady state T feed could result several T outlet. Hysteresis phenomenon may result asset losses like explosion, leakage, and loosing material integrity. Double bed reactor configuration allows us to use several operating parameters as variation to overcome hysteresis. In this review, quenching ratio was chosen to be that varied parameter. This study aims to determine how quenching ratio affects hysteresis zone by utilizing Aspen Hysys® V8.8 as simulation tool. Simulation showed that quenching ratio would narrow hysteresis zone yet increased extinction temperature that lower the conversion. Conversion profile showed that 0.2 quenching ratio got the highest conversion for system with bed volume ratio 2:1 while total volume was 30 m3. However, the feed temperature was fallen at hysteresis zone. Dynamic simulation showed that highest conversion feed temperature (10%ΔTf above extinct temperature was still able to preserve stability with descending temperature approach. Hysteresis itself started to occur at 1.7%ΔTf above extinct temperature
Energy Technology Data Exchange (ETDEWEB)
Cuperman, S. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel); Bruma, C. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel) and College of Judea and Samaria, 44837 Ariel (Israel)]. E-mail: edycb@post.tau.ac.il; Komoshvili, K. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel); College of Judea and Samaria, 44837 Ariel (Israel)
2007-03-05
The combined plasma non-uniformity effects on the energy deposition of Alfven waves launched by an external antenna in pre-heated spherical tokamaks are investigated. The following relevant physical processes are here possible: (a) the emergence of gaps in the shear Alfven continuum spectrum and the generation of discrete global Alfven eigenmodes with frequencies inside the gaps; (b) multi-wave interactions, interactions of gaps of the same kind (e.g., toroidicity induced) and of different kinds (toroidicity, ellipticity and triangularity induced) as well as of secondary order gaps arising when a pair of modes is coupled to one or more modes through other coupling parameters; (c) basic wave-plasma interactions as propagation, reflection, mode-conversion, tunneling and deposition. Thus, we solved numerically the full 2D wave equations for the vector and scalar potentials, using a quite general two-fluid resistive tensor-operator, without any geometrical limitations. The results obtained indicate the existence of antenna-launched wave characteristics for which the power is most efficiently coupled in outer regions of plasmas, which is of special interest for low aspect ratio tokamaks, e.g., for the generation of non-inductive current drive as well as for turbulence suppression and transport barriers formation.
Fusion potential for spherical and compact tokamaks
International Nuclear Information System (INIS)
Sandzelius, Mikael
2003-02-01
The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect
Fusion potential for spherical and compact tokamaks
Energy Technology Data Exchange (ETDEWEB)
Sandzelius, Mikael
2003-02-01
The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.
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)
International Nuclear Information System (INIS)
Bruma, C; Cuperman, S; Komoshvili, K
2003-01-01
In an effort to optimize the internal transport barriers (ITBs) generated by externally launched mode-converted fast waves (FWs) in pre-heated spherical tokamaks (STs), we have carried out a systematic parametric investigation with respect to the rf waves and antenna characteristics; as a study case, a START-like device has been considered. Within the framework of a plasma model including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped and passing electrons and ions, and starting with the solution of the full wave equation for a ST-plasma, we show that optimized ITBs, suitable for the stabilization of plasma turbulence (e.g. overpassing the maximum growth rate of the ITG-instability, γ ITG ) in STs can be generated by the aid of externally launched FW and mode-converted to kinetic Alfven waves. This result holds in spite of the limiting trapped-particles associated squeezing factor S present in the non-linear equation for E r (via the viscosity coefficient μ θi ∝|S| 3/2 , S = S(dE r /dr))
Energy Technology Data Exchange (ETDEWEB)
Bruma, C [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Cuperman, S [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Komoshvili, K [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel)
2003-04-01
In an effort to optimize the internal transport barriers (ITBs) generated by externally launched mode-converted fast waves (FWs) in pre-heated spherical tokamaks (STs), we have carried out a systematic parametric investigation with respect to the rf waves and antenna characteristics; as a study case, a START-like device has been considered. Within the framework of a plasma model including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped and passing electrons and ions, and starting with the solution of the full wave equation for a ST-plasma, we show that optimized ITBs, suitable for the stabilization of plasma turbulence (e.g. overpassing the maximum growth rate of the ITG-instability, {gamma}{sub ITG}) in STs can be generated by the aid of externally launched FW and mode-converted to kinetic Alfven waves. This result holds in spite of the limiting trapped-particles associated squeezing factor S present in the non-linear equation for E{sub r} (via the viscosity coefficient {mu}{sub {theta}}{sub i}{proportional_to}|S|{sup 3/2}, S = S(dE{sub r}/dr))
Sheffield, J.
1981-08-01
For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.
The Ar{sup 17+} Ly{sub {alpha}2}/Ly{sub {alpha}1} ratio in Alcator C-Mod tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Rice, J E; Reinke, M L; Ince-Cushman, A C; Podpaly, Y A [Plasma Science and Fusion Center, MIT, Cambridge, MA (United States); Ashbourn, J M A [Mathematical Institute, University of Oxford, Oxford (United Kingdom); Gu, M F [SSL, University of California Berkeley, CA (United States); Bitter, M; Hill, K [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Rachlew, E, E-mail: rice@psfc.mit.edu [KTH, Stockholm (Sweden)
2011-08-28
High-quality spectra of hydrogen-like Ar{sup 17+} have been obtained from Alcator C-Mod tokamak plasmas using a spatially imaging high-resolution x-ray spectrometer system in an extensive study of the underlying high-n satellite lines. The ratio of Ly{sub {alpha}2} (1S{sub 1/2}-2P{sub 1/2}) to Ly{sub {alpha}1} (1S{sub 1/2}-2P{sub 3/2}) was found to be {approx}0.52 regardless of plasma parameters, which is somewhat greater than the ratio of the statistical weights of the upper n = 2 levels, 0.5. This difference is mainly due to the effects of collisional excitation of fine-structure sub-levels. For the observations presented here, electron densities were in an extended range from 3x10{sup 19} to 4x10{sup 20} m{sup -3} with electron and ion temperatures between 1 and 4 keV. Experimental results are compared to calculations from COLRAD, a collisional-radiative modelling code, and good agreement is shown.
Effects of magnetic shear on current penetration in a tokamak
International Nuclear Information System (INIS)
Zhang Pengyun; Wang Long
2001-01-01
The penetrations of the parallel and perpendicular components of plasma currents are interrelated to each other due to the existence of magnetic shear in a tokamak configuration. Effects of the shear on the penetration of Fourier components of toroidal plasma current are analysed in a cylindrical column model. The current penetration is obviously strengthened by the shear for a bell-bike conductivity profile and low safety factor and low aspect ratio
International Nuclear Information System (INIS)
1991-05-01
The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway
START: the creation of a spherical tokamak
International Nuclear Information System (INIS)
Sykes, Alan
1992-01-01
The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)
Energy Technology Data Exchange (ETDEWEB)
Cuperman, S.; Bruma, C.; Komoshvili, K
2003-05-12
The generation in low aspect ratio tokamaks (LARTs) of ponderomotive forces and non-inductive current drive by the resonant fast wave-plasma interaction with mode conversion to kinetic Alfven waves (KAWs) and subsequent deposition, mainly by resonant electron Landau damping, is considered. The calculations follow the rigorous solution of the full wave equations upon using a dielectric tensor operator consisting of (i) a parallel conductivity including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped electrons and passing electrons+ions and (ii) perpendicular components provided by the resistive two-fluid model equations. The fast waves are launched by an antenna located on the low field side and extending {+-}45 deg. about the equatorial plane. A parametric investigation of the structure and importance of the various components of the ponderomotive forces and current drive generated in START-like plasmas is carried out and their suitability for supplementing the required non-rf toroidal equilibrium current is demonstrated.
International Nuclear Information System (INIS)
Berger, D.; Bernard, L.C.; Gruber, R.; Troyon, F.
1980-01-01
The Lausanne ideal MHD stability code ERATO is used to investigate spectral properties of Solovev's equilibrium at small aspect ratios. Two different elongations are considered. Both free and rigid boundary models are computed and compared. Modes characterized by a large radial extension have been found which appear to be due to coupling of m=1 and m=2 modes due to toricity. The internal modes spectrum is compared with the predictions of the full Mercier criterion, taking into account its spatial dependence, and with the ballooning modes stability criterion. (Auth.)
The spheric tokamak programme at Culham
International Nuclear Information System (INIS)
Sykes, A.
1999-01-01
The Spherical Tokamak (ST) is the low aspect ratio limit of the conventional tokamak, and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration, and was operational at Culham from January 1991 to March 1998, obtaining plasma current of up to 300 kA and pulse durations of ∼ 50 ms. Its successor, MAST is scheduled to obtain first plasma in Autumn 1998 and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high-performance central solenoid, 1.5 MW ECRH and 5 MW of Neutral Beam Injection. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose-built STs (such as MAST) are described. (author)
Advanced commercial tokamak study
International Nuclear Information System (INIS)
Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.
1985-12-01
Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs
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
Combined confinement system applied to tokamaks
International Nuclear Information System (INIS)
Ohkawa, Tihiro
1986-01-01
From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)
Numerical study of spherical Torus MHD equilibrium configuration
International Nuclear Information System (INIS)
Cheng Faying; Dong Jiaqi; Wang Aike
2003-01-01
Tokamak equilibrium code SWEQU has been modified so that it can be used for the MHD equilibrium study of low aspect ratio device. Evolution of plasma configuration in start-up phase and double-null divertor configuration in steady-state phase has been simulated using the modified code. Results show that the new code can be used not only to obtain the equilibrium configuration of spherical Torus in steady-state phase, but also to simulate the evolution of plasma in the start-up phase
New directions in tokamak reactors
International Nuclear Information System (INIS)
Baker, C.C.
1985-01-01
New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
1999-01-01
Theoretical results on the wave-plasma interactions in simulated toroidal configurations are presented. The study covers the cases of large to low aspect ratio tokamaks, in the pre-heated stage. Fast waves emitted from an external antenna with different wave numbers and frequencies are considered. The non-inductive Alfven wave current drive is evaluated and discussed. (author)
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
2001-01-01
Theoretical results on the wave-plasma interactions in simulated toroidal configurations are presented. The study covers the cases of large to low aspect ratio tokamaks, in the pre-heated stage. Fast waves emitted from an external antenna with different wave numbers and frequencies are considered. The non-inductive Alfven wave current drive is evaluated and discussed. (author)
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
Engineering Design of KSTAR tokamak main structure
International Nuclear Information System (INIS)
Im, K.H.; Cho, S.; Her, N.I.
2001-01-01
The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)
Brandon, Jay M.; Foster, John V.; Shah, Gautam H.; Gato, William; Wilborn, James E.
2004-01-01
Improvements in testing and modeling of nonlinear and unsteady aerodynamic effects for flight dynamics predictions of vehicle performance is critical to enable the design and implementation of new, innovative vehicle concepts. Any configuration which exhibits significant flow separation, nonlinear aerodynamics, control interactions or attempts maneuvering through one or more conditions such as these is, at present, a challenge to test, model or predict flight dynamic responses prior to flight. Even in flight test experiments, adequate models are not available to study and characterize the complex nonlinear and time-dependent flow effects occurring during portions of the maneuvering envelope. Traditionally, airplane designs have been conducted to avoid these areas of the flight envelope. Better understanding and characterization of these flight regimes may not only reduce risk and cost of flight test development programs, but also may pave the way for exploitation of those characteristics that increase airplane capabilities. One of the hurdles is that the nonlinear/unsteady effects appear to be configuration dependent. This paper compares some of the dynamic aerodynamic stability characteristics of two very different configurations - representative of a fighter and a transport airplane - during dynamic body-axis roll wind tunnel tests. The fighter model shows significant effects of oscillation frequency which are not as apparent for the transport configuration.
International Nuclear Information System (INIS)
Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.
1997-01-01
The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs
Three-dimensional tokamak equilibria and stellarators with two-dimensional magnetic symmetry
International Nuclear Information System (INIS)
Garabedian, P.R.
1997-01-01
Three-dimensional computer codes have been developed to simulate equilibrium, stability and transport in tokamaks and stellarators. Bifurcated solutions of the tokamak problem suggest that three-dimensional effects may be more important than has generally been thought. Extensive calculations have led to the discovery of a stellarator configuration with just two field periods and with aspect ratio 3.2 that has a magnetic field spectrum B mn with toroidal symmetry. Numerical studies of equilibrium, stability and transport for this new device, called the Modular Helias-like Heliac 2 (MHH2), will be presented. (author)
International Nuclear Information System (INIS)
1991-01-01
The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs
Shimizu, Yukimaru; Sugino, Koichi; Yasui, Masaji; Hayakawa, Yukitaka; Kuzuhara, Sadao
1985-01-01
Pipes with bend combinations are much used in the heat exchangers, since the curved path in the bends promotes the mixing in flow for active heat transfer. In the present paper, one of the pipes with bend combinations, namely, quasi-coiled pipes composed of many bend elements are investigated, and the relationships between the hydraulic loss and the secondary flow are studied experimentally. The configurations of the cross sections, the bent angles and the curvature ratios of the bend element...
MHD stability of an almost circular tokamak
International Nuclear Information System (INIS)
Roy, A.
1990-10-01
In a tokamak, the ratio β between the plasma pressure and that of the magnetic field is limited by the appearance of instabilities. The magnetic field in a tokamak reactor will always be limited by technological constraints. It is therefore crucial to know what factors have an effect on the β limit, since a zero resistivity plasma fluid model allows for theoretical reproduction of the β limits observed experimentally. Theoretical studies have shown that the distributions of pressure and current density may have a substantial effect on the β limit. The effect of the current density and pressure distributions on the β limit has been studied for tokamak with a circular core section. The best results are obtained when the current density is concentrated in the centre of the section and is nil at the periphery. But the second region of stability against ballooning modes cannot be obtained in a circular tokamak owing to the destabilisation of the universal modes. This study was then extended to the stability of plasmas the section of which is almost circular and has a point of reflection. Such configurations are vital for fusion since they allow systems in which the confinement time does not deteriorate with an increase in the additional heating power. The β limit was calculated for different positions of the reflection point. The results show that when it is displaced from the interior towards the exterior of the torus, the stability of the overall modes is progressively improved until it is vertical. But if the point of reflection is further displaced from this vertical position towards the exterior of the torus, localised modes close to the edge of the plasma are destabilised and bring about a drop in the β limit. (author) figs., tabs., 80 refs
International Nuclear Information System (INIS)
Francis Maria Anand, L.; Gudennavar, S.B.; Bubbly, S.G.; Joseph, Daisy
2013-01-01
The article presents a simple method of measuring K-shell absorption jump ratios and jump factors for elements in the field of X-ray spectroscopy. The K-shell jump ratios and jump factors for Molybdenum and Silver are measured by adopting 2ŏ-geometrical configuration and a weak gamma source. The characteristic K X-ray photons are excited in the targets using 32.8 keV barium X-ray photons from a weak 137 Cs radioactive source that is produced due to the internal conversion of cesium nucleus (IC). The fluorescent K X-ray photons are detected using low energy Si(Li) detector coupled to a 8k multichannel analyser. The K X-ray intensity ratios from X-ray fluorescent spectrum are measured experimentally, the total atomic attenuation cross section and the total atomic scattering cross sections are calculated using WinXcom software. The K-shell jump factor and jump ratio are computed using the measured K X-ray intensity ratios and the calculated K a , X-ray production cross section. The computed values of K-shell jump factor and jump ratio for molybdenum and silver are compared with the theoretical values and others' experimental data and are presented. The amount of uncertainty in the experimental measurement of K X-ray intensity ratios is less than 5%. Thus the 2ŏ-geometrical configuration method with weak gamma source can be an alternative simple method to measure the jump factors and the jump ratios of pure elements in the field of X-ray spectroscopy. (author)
International Nuclear Information System (INIS)
Dini, F.; Khorasani, S.
2007-01-01
Configuration of Tokamak plasma has a dominant effect on its parameters. In the calculation of transport, there are some transport coefficients and quantities, where the knowledge of their precise values, according to the system of equations, is essential to be realized. Tokamak has a toroidal configuration, in addition to classical effects, it is necessary to study the neoclassical effects due to the field curvature. The trapped particles in strong electromagnetic fields oscillate on banana-shaped orbits which in turn affect many other collisional transport parameters. Here, a precise estimation of trapped particles based on the standard equilibrium model for an elliptical shape of Tokamak plasma has been carried out using Lin-Liu model. It should be added that in this calculation, the profile of the averaged magnetic field on the flux surfaces has been derived using analytical integration and consideration of an elliptic shape for ellipticity function in the limit of large aspect ratio and zero shift of magnetic flux surfaces. Having the fraction of the trapped particles, by ,following the formulation and using an appropriate model in various collisional regimes, the neoclassical conductivity of plasma in Damavand Tokamak is obtained and the respective variations have been found. The presented results can exploit the computation of transport and other quantities of Damavand Tokamak
International Nuclear Information System (INIS)
Weldon, D.M.; Murray, J.G.
1983-01-01
Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor
Tokamak fusion reactor exhaust
International Nuclear Information System (INIS)
Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.
1981-08-01
This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)
Dewhurst, J.; Hnat, B.; Dudson, B.; Dendy, R. O.; Counsell, G. F.; Kirk, A.
2007-12-01
Almost all astrophysical and magnetically confined fusion plasmas are turbulent. Here, we examine ion saturation current (Isat) measurements of edge plasma turbulence for three MAST L-mode plasmas that differ primarily in their edge magnetic field configurations. First, absolute moments of the coarse grained data are examined to obtain accurate values of scaling exponents. The dual scaling behaviour is identified in all samples, with the temporal scale τ ≍ 40-60 μs separating the two regimes. Strong universality is then identified in the functional form of the probability density function (PDF) for Isat fluctuations, which is well approximated by the Fréchet distribution on temporal scales τ ≤ 40μs. For temporal scales τ > 40μs, the PDFs appear to converge to the Gumbel distribution, which has been previously identified as a universal feature of many other complex phenomena. The optimal fitting parameters k=1.15 for Fréchet and a=1.35 for Gumbel provide a simple quantitative characterisation of the full spectrum of fluctuations. We conclude that, to good approximation, the properties of the edge turbulence are independent of the edge magnetic field configuration.
Hnat, B.; Dudson, B. D.; Dendy, R. O.; Counsell, G. F.; Kirk, A.; MAST Team
2008-08-01
Ion saturation current (Isat) measurements of edge plasma turbulence are analysed for six MAST L-mode plasmas that differ primarily in their edge magnetic field configurations. The analysis techniques are designed to capture the strong nonlinearities of the datasets. First, absolute moments of the data are examined to obtain accurate values of scaling exponents. This confirms dual scaling behaviour in all samples, with the temporal scale τ ≈ 40-60 µs separating the two regimes. Strong universality is then identified in the functional form of the probability density function (PDF) for Isat fluctuations, which is well approximated by the Fréchet distribution on temporal scales τ 40 µs, the PDFs appear to converge to the Gumbel distribution, which has been previously identified as a universal feature of many other complex phenomena. The optimal fitting parameters k = 1.15 for Fréchet and a = 1.35 for Gumbel provide a simple quantitative characterization of the full spectrum of fluctuations. It is concluded that, to good approximation, the properties of the edge turbulence are independent of the edge magnetic field configuration.
Study of a compact reversed shear Tokamak reactor
International Nuclear Information System (INIS)
Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.
1998-01-01
A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)
Modular coils and finite-β operation of a quasi-axially symmetric tokamak
International Nuclear Information System (INIS)
Drevlak, M.
1998-01-01
Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nuehrenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (MERKEL, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (HIRSHMAN, S.P., VAN RIJ, W.I., MERKEL, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak
Han, I.; Demir, L.
2009-11-01
Kβ -to- Kα x-ray intensity ratios of Fe, Cr, and Ni have been measured in pure metals and in alloys of FexNi1-x ( x=0.8 , 0.7, 0.6, 0.5, 0.4, 0.3, and 0.2), NixCr1-x ( x=0.8 , 0.6, 0.5, 0.4, and 0.2), FexCr1-x ( x=0.9 , 0.7, and 0.5), and FexCryNi1-(x+y) ( x=0.7-y=0.1 , x=0.5-y=0.2 , x=0.4-y=0.3 , x=0.3-y=0.3 , x=0.2-y=0.2 , and x=0.1-y=0.2 ) following excitation by 22.69 keV x rays from a 10 mCi C109d radioactive point source. The valence-electron configurations of these metals were determined by corporation of measured Kβ -to- Kα x-ray intensity ratios with the results of multiconfiguration Dirac-Fock calculation for various valence-electron configurations. Valence-electron configurations of 3d transition metals in alloys indicate significant differences with respect to the pure metals. Our analysis indicates that these differences arise from delocalization and/or charge transfer phenomena in alloys. Namely, the observed change of the valence-electron configurations of metals in alloys can be explained with the transfer of 3d electrons from one element to the other element and/or the rearrangement of electrons between 3d and 4s,4p states of individual metal atoms.
Demonstration tokamak power plant
International Nuclear Information System (INIS)
Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.
1983-01-01
A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System
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
MAST: a Mega Amp Spherical Tokamak
International Nuclear Information System (INIS)
Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.
1995-01-01
The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)
Instrumentation and controls of an ignited tokamak
International Nuclear Information System (INIS)
Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega, R.J.; Raju, G.V.S.; Stone, R.S.
1980-10-01
The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented
Full power in the European tokamak
International Nuclear Information System (INIS)
Lallia, P.P.; Hugon, M.
1987-01-01
A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years
Tokamak power systems studies at ANL
International Nuclear Information System (INIS)
Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.
1986-01-01
A number of advances in plasma physics and engineering promise to greatly improve the reactor prospects of tokamaks. The following features, in particular, are examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 7 T; (c) low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields produced in the plasma. In addition to matching desirable high-beta equilibria, this method is capable of producing a large variety of new equilibria, many of which look attractive. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors
International Nuclear Information System (INIS)
Lee, Young Hyun; Lee, Jung Kun; Hong, Kug Sun
2003-01-01
The (100) and (111) oriented Pb(Zr 1-x Ti x )O 3 (PZT) thins films were prepared on Pt/TiO 2 /SiO 2 /Si(100) substrate using sol-gel process. The ferroelectric and dielectric properties were evaluated as a function of Zr/Ti ratio and the origin of their changes were discussed in terms of domain engineering. With increasing Zr/Ti ratio, the remnant polarization of the polar axis decreased from 23.87 to 10.66 μC/cm 2 and the difference of the saturated polarization between polar axis and non-polar axis increased systematically. The fatigue rate of PZT films was disproportional to Zr/Ti ratio. At a composition of x = 0.7, the remnant polarization decreased from 2P r = 30.82 μC/cm 2 to 26.60 μC/cm 2 after 6.9 x 10 10 cycles. The fatigue rate was also related to the direction of polarization reversal and the increased fatigue behavior was observed in polar axis. These results were discussed with the aid of the domain configuration and the switching current, which revealed the role of the internal stress on ferroelectric property of PZT films
Stability of high-beta tokamak equilibria and transport in Belt-Pinch IIa
Energy Technology Data Exchange (ETDEWEB)
Becker, G; Gruber, O; Krause, H; Mast, F; Wilhelm, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.)
1978-01-01
In Belt-Pinch IIa, highly elongated equilibria with poloidal beta values up to the aspect ratio have been achieved. In these tokamak-like configurations, no fast-growing MHD instabilities such as external kink and ballooning modes have been observed. Rigid displacement instabilities have been stabilized by an appropriate poloidal magnetic field configuration and by a conducting shell. By comparing simulation experiments using the Garching high-beta transport code with measurements, it has been found that in the collision-dominated plasma no anomalously enhanced transport occurs. Transport theory in the Pfirsch-Schlueter regime, which includes elongation and high-beta effects, has been confirmed by the experiment. In particular, it has been shown that the perpendicular electrical conductivity is also classical. Detailed investigations of oxygen and carbon impurity losses demonstrated that the impurity subprograms commonly used for tokamaks underestimate the radiation losses in the range Tsub(e)=10 to 30 eV.
Tokamak residual zonal flow level in near-separatrix region
International Nuclear Information System (INIS)
Bing-Ren, Shi
2010-01-01
Residual zonal flow level is calculated for tokamak plasmas in the near-separatrix region of a diverted tokamak. A recently developed method is used to construct an analytic divertor tokamak configuration. It is shown that the residual zonal flow level becomes smaller but still keeps finite near the separatrix because the neoclassical polarisation mostly due to the trapped particles goes larger in this region. (fluids, plasmas and electric discharges)
International Nuclear Information System (INIS)
Khakoo, M. A.; Wrkich, J.; Larsen, M.; Kleiban, G.; Kanik, I.; Trajmar, S.; Brunger, M.J.; Teubner, P.J.O.; Crowe, A.; Fontes, C.J.; Clark, R.E.H.; Zeman, V.; Bartschat, K.; Madison, D.H.; Srivastava, R.; Stauffer, A.D.
2002-01-01
Electron-impact differential cross-section measurements for the excitation of the 2p 5 3s configuration of Ne are reported. The Ne cross sections are obtained using experimental differential cross sections for the electron-impact excitation of the n=2 levels of atomic hydrogen [Khakoo et al., Phys. Rev. A 61, 012701-1 (1999)], and existing experimental helium differential cross-section measurements, as calibration standards. These calibration measurements were made using the method of gas mixtures (Ne and H followed by Ne and He), in which the gas beam profiles of the mixed gases are found to be the same within our experimental errors. We also present results from calculations of these differential cross sections using the R-matrix and unitarized first-order many-body theory, the distorted-wave Born approximation, and relativistic distorted-wave methods. Comparison with available experimental differential cross sections and differential cross-section ratios is also presented
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)
Non-Axisymmetric Shaping of Tokamaks Preserving Quasi-Axisymmetry
Energy Technology Data Exchange (ETDEWEB)
Long-Poe Ku and Allen H. Boozer
2009-06-05
If quasi-axisymmetry is preserved, non-axisymmetric shaping can be used to design tokamaks that do not require current drive, are resilient to disruptions, and have robust plasma stability without feedback. Suggestions for addressing the critical issues of tokamaks can only be validated when presented with sufficient specificity that validating experiments can be designed. The purpose of this paper is provide that specificity for non-axisymmetric shaping. To our knowledge, no other suggestions for the solution of a number of tokamak issues, such as disruptions, have reached this level of specificity. Sequences of three-field-period quasi-axisymmetric plasmas are studied. These sequences address the questions: (1) What can be achieved at various levels of non-axisymmetric shaping? (2) What simplifications to the coils can be achieved by going to a larger aspect ratio? (3) What range of shaping can be achieved in a single experimental facility? The sequences of plasmas found in this study provide a set of interesting and potentially important configurations.
International Nuclear Information System (INIS)
Cumyn, P.B.
A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics
International Nuclear Information System (INIS)
Jarboe, T.R.; Nelson, B.A.
1997-01-01
During the reporting period of this HIT grant (1 June 1995--31 May 1997) we've conducted further stability analysis, used the TIP diagnostic to measure plasma fields in HIT, and developed a single-parameter helicity injector model. HIT has undergone a significant upgrade to the HIT-II configuration which is described here. Parts for HIT-II have been designed, ordered, and received under this grant and are being assembled under the subsequent grant
Heavy ion beam probe development for the plasma potential measurement on the TUMAN-3M tokamak
International Nuclear Information System (INIS)
Askinazi, L.G.; Kornev, V.A.; Lebedev, S.V.; Tukachinsky, A.S.; Zhubr, N.A.; Dreval, N.B.; Krupnik, L.I.
2004-01-01
The peculiarities of the heavy ion beam probe implementation on the small aspect ratio tokamak TUMAN-3M are analyzed. The toroidal displacement of beam trajectory due to the high I pl /B tor ratio is taken into account when designing the layout of the diagnostic. Numerical calculation of beam trajectories using realistic configuration of TUMAN-3M magnetic fields and parabolic plasma current profile resulted in proper adjustment of probing and detection parameters (probing ion material, energy, entrance angles, detector location, and orientation). Secondary ion energy analyzer gain functions G and F were measured in situ using neutral hydrogen puffed in the tokamak vessel as a target for secondary ions production. The detector unit featured split-plate design and had additional electrodes for secondary electron emission suppression. As a result, the diagnostic is now capable of plasma potential evolution measurement and is sensitive enough to trace the potential profile evolution at the L-H mode transition
Axisymmetric control in tokamaks
International Nuclear Information System (INIS)
Humphreys, D.A.
1991-02-01
Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration
Prospects for Tokamak Fusion Reactors
International Nuclear Information System (INIS)
Sheffield, J.; Galambos, J.
1995-01-01
This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant
Integral torque balance in tokamaks
International Nuclear Information System (INIS)
Pustovitov, V.D.
2011-01-01
The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.
Equilibrium: two-dimensional configurations
International Nuclear Information System (INIS)
Anon.
1987-01-01
In Chapter 6, the problem of toroidal force balance is addressed in the simplest, nontrivial two-dimensional geometry, that of an axisymmetric torus. A derivation is presented of the Grad-Shafranov equation, the basic equation describing axisymmetric toroidal equilibrium. The solutions to equations provide a complete description of ideal MHD equilibria: radial pressure balance, toroidal force balance, equilibrium Beta limits, rotational transform, shear, magnetic wall, etc. A wide number of configurations are accurately modeled by the Grad-Shafranov equation. Among them are all types of tokamaks, the spheromak, the reversed field pinch, and toroidal multipoles. An important aspect of the analysis is the use of asymptotic expansions, with an inverse aspect ratio serving as the expansion parameter. In addition, an equation similar to the Grad-Shafranov equation, but for helically symmetric equilibria, is presented. This equation represents the leading-order description low-Beta and high-Beta stellarators, heliacs, and the Elmo bumpy torus. The solutions all correspond to infinitely long straight helices. Bending such a configuration into a torus requires a full three-dimensional calculation and is discussed in Chapter 7
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)
Development of Operation Scenario for Spherical Tokamak at SNU
International Nuclear Information System (INIS)
Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.
2009-01-01
Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed
Flux surface shaping effects on tokamak edge turbulence and flows
International Nuclear Information System (INIS)
Kendl, A.; Scott, B.D.
2004-01-01
The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 ≤ κ ≥ 2 and triangularity 0 ≤ δ ≤ 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)
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)
Flux surface shaping effects on tokamak edge turbulence and flows
Energy Technology Data Exchange (ETDEWEB)
Kendl, A. [Innsbruck Univ., Institut fuer Theoretische Physik, Association EURATOM (Austria); Scott, B.D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching bei Muenchen (Germany)
2004-07-01
The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 {<=} {kappa} {>=} 2 and triangularity 0 {<=} {delta} {<=} 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)
Physics Issues in the Design of Low Aspect-Ratio, High-Beta, Quasi-Axisymmetric Stellarators
International Nuclear Information System (INIS)
Zarnstorff, M.C.; Berry, L.A.; Boozer, A.; Brooks, A.; Cooper, W.A.
2000-01-01
Compact stellarators have the potential to combine the best features of the stellarator and the advanced tokamak, offering steady state operation without current drive and potentially without disruptions at an aspect ratio similar to tokamaks. A quasi-axisymmetric stellarator is developed that is consistent with the boot-strap current and passively stable to the ballooning, kink, Mercier, vertical, and neoclassical tearing modes at b=4.1 % without need for conducting walls or external feedback. The configuration has good flux surfaces and fast ion confinement. Thermal transport analysis indicates that the confinement should be similar to tokamaks of the same size, allowing access to the b-limit with moderate power. Coils have been designed to reproduce the physics properties. Initial analysis indicates the coils have considerable flexibility to manipulate the configuration properties. Simulations of the current evolution indicate the kink-mode can remain stable during the approach to h igh-beta
Improvement of the tokamak concept
Energy Technology Data Exchange (ETDEWEB)
Laurent, L
1994-12-31
Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.
Comprehensive numerical modelling of tokamaks
International Nuclear Information System (INIS)
Cohen, R.H.; Cohen, B.I.; Dubois, P.F.
1991-01-01
We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell
Computation of tokamak equilibria with steady flow
International Nuclear Information System (INIS)
Kerner, W.; Tokuda, Shinji
1987-08-01
The equations for ideal MHD equilibria with stationary flow are reexamined and addressed as numerically applied to tokamak configurations with a free plasma boundary. Both the isothermal (purely toroidal flow) and the poloidal flow cases are treated. Experiment-relevant states with steady flow (so far only in the toroidal direction) are computed by the modified SELENE40 code. (author)
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
Modelling ohmic confinement experiments on the START tokamak
International Nuclear Information System (INIS)
Roach, C.M.
1996-05-01
Ohmic confinement data from the tight aspect ratio tokamak START has been analysed using the ASTRA transport simulation code. Neoclassical expressions have been modified to describe tight aspect ratio configurations, and the comparison between START data and models of anomalous transport has been made quantitative using the standard χ 2 test from statistics. Four confinement models (T11, Rebut-Lallia-Watkins, Lackner-Gottardi, and Taroni et al's Bohm model) have been compared with the START data. Three of the models are found to simulate START's electron temperature data moderately well, while Taroni et al's Bohm model overestimates electron temperatures in START by an order of magnitude. Thus comparison with START data tends to discriminate against Bohm models; these models are pessimistic or ITER. (author)
International Nuclear Information System (INIS)
Rawls, J.M.
1979-10-01
An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design
Tokamak and RFP ignition requirements
International Nuclear Information System (INIS)
Werley, K.A.
1991-01-01
A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab
International Nuclear Information System (INIS)
Jarboe, T.R.; Nelson, B.A.
1995-01-01
This is the fifth Progress Report on the Helicity Injected Tokamak (HIT) at the University of Washington, Seattle, DOE Grant DE-FE06-90ER54095. This report covers the period of December 1, 1993 through December 31
Thermonuclear ignition in the next generation tokamaks
International Nuclear Information System (INIS)
Johner, J.
1989-04-01
The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aB t x of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition
Tokamak devices: towards controlled fusion
International Nuclear Information System (INIS)
Trocheris, M.
1975-01-01
The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr
Extremely shaped plasmas to improve the Tokamak concept
Energy Technology Data Exchange (ETDEWEB)
Piras, F.
2011-04-15
experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is
Extremely shaped plasmas to improve the Tokamak concept
International Nuclear Information System (INIS)
Piras, F.
2011-04-01
experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is
Analysis of equilibrium in a tokamak by the finite-difference method
International Nuclear Information System (INIS)
Kim, K.E.; Jeun, G.D.
1983-01-01
Ideal magnetohydrodynamic equilibrium in a Tokamak having a small radius with an elongated rectangular cross section is studied by applying the finite-difference method to the Grad-Shafranov equation to determine possible limitations for *b=8*pPsup(2)/Bsup(2). The coupled first-order differential equations resulting from the finite-difference Grad-Shafranov equation is solved by the numarical method:1)We concluded that equilibrium consideration alone gives no limitation even for *b approx.1. 2)We have obtained the equilibrium magnetic field configuration charcterized by a set of three parameters;the aspect ratio, *b,and the safety factor. (Author)
International Nuclear Information System (INIS)
Song, Yuntao; Wu, Weiyue; Du, Shijun
2014-01-01
Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.
International Nuclear Information System (INIS)
Nishio, Satoshi
1998-01-01
If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)
Tokamak fusion reactors with less than full tritium breeding
International Nuclear Information System (INIS)
Evans, K. Jr.; Gilligan, J.G.; Jung, J.
1983-05-01
A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed
MHD stability properties of bean-shaped tokamaks
International Nuclear Information System (INIS)
Grimm, R.C.; Chance, M.S.; Todd, A.M.M.
1984-03-01
A study of the MHD stability properties of bean-shaped tokamak plasmas is presented. For ballooning modes, while increased indentation gives larger β stable configurations, the existence and accessibility of the second stable region is sensitive to the pressure and safety factor profiles. The second stable region appears at lower β values for large aspect ratio and moderately high q-values. Finite-Larmor-radius (FLR) kinetic effects can significantly improve the stability properties. For low q (< 1) operation, long wavelength (n approx. 2,3) internal pressure driven modes occur at modest β/sub p/ values and accessibility to higher β operation is unlikely. Indentation modifies the nature of the usually vertical axisymmetric instability, but the mode can be passively stabilized by placing highly conducting plates near to the tips of the plasma bean. At constant q, indentation has a stabilizing effect on tearing modes
Drift-kinetic Alfven modes in high performance tokamaks
International Nuclear Information System (INIS)
Jaun, A.; Fasoli, A.F.; Testa, D.; Vaclavik, J.; Villard, L.
2001-01-01
The stability of fast-particle driven Alfven eigenmodes is modeled in high performance tokamaks, successively with a conventional shear, an optimized shear and a tight aspect ratio plasma. A large bulk pressure yields global kinetic Alfven eigenmodes that are stabilized by mode conversion in the presence of a divertor. This suggests how conventional reactor scenarii could withstand significant pressure gradients from the fusion products. A large safety factor in the core q 0 >2.5 in deeply shear reversed configurations and a relatively large bulk ion Larmor radius in a low magnetic field can trigger global drift-kinetic Alfven eigenmodes that are unstable in high performance JET, NSTX and ITER plasmas. (author)
Song, Yuntao; Du, Shijun
2013-01-01
Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study
Advanced Tokamak Stability Theory
Zheng, Linjin
2015-03-01
The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.
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
19 rectifiers to supply the coils of the TCV tokamak
International Nuclear Information System (INIS)
Fasel, D.; Perez, A.; Depreville, G.; Puchar, F.; Pahud, J.D.
1990-01-01
This paper describes the electrical network designed to supply the 19 coils of the TCV (Tokamak a Configuration Variable) tokamak. After a brief description of the main purpose of TCV, the general characteristics of the TCV network are given. Then the technical choices made for the rectifier power stage are detailed. There follows a description of the rectifier digital control electronics. Comments on simulations carried out and the actual status conclude the paper. (author) 3 refs., 5 figs., 2 tabs
/sup 3/He functions in tokamak-pumped laser systems
Energy Technology Data Exchange (ETDEWEB)
Jassby, D.L.
1986-10-01
/sup 3/He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the /sup 3/He(n,p)T reaction, and thereby excite gaseous lasants mixed with the /sup 3/He while simultaneously breeding tritium. The total /sup 3/He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak.
3He functions in tokamak-pumped laser systems
International Nuclear Information System (INIS)
Jassby, D.L.
1986-10-01
3 He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the 3 He(n,p)T reaction, and thereby excite gaseous lasants mixed with the 3 He while simultaneously breeding tritium. The total 3 He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak
A computer code for Tokamak reactor concepts evaluation
International Nuclear Information System (INIS)
Rosatelli, F.; Raia, G.
1985-01-01
A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts
Tokamak confinement scaling laws
International Nuclear Information System (INIS)
Connor, J.
1998-01-01
The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)
International Nuclear Information System (INIS)
1986-04-01
This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking
Tokamak power system studies at ANL
International Nuclear Information System (INIS)
Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.
1986-06-01
The following features, in particular, have been examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I ≅ 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors
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
Tokamak Fusion Core Experiment maintenance study
International Nuclear Information System (INIS)
Snyder, A.M.; Watts, K.D.
1985-01-01
The recently completed Tokamak Fusion Core Experiment (TFCX) design project was carried out to investigate potential next generation tokamak concepts. An important aspect of this project was the early development and incorporation of remote maintainability throughout the design process. This early coordination and incorporation of maintenance aspects to the design of the device and facilities would assure that the machine could ultimately be maintained and repaired in an efficient and cost effective manner. To meet this end, a rigorously formatted engineering trade study was performed to determine the preferred configuration for the TFCX reactor based primarily on maintenance requirements. The study indicated that the preferred design was one with an external vacuum vessel and torrodial field coils that could be removed via a simple radial motion. The trade study is presented and the preferred TFCX configuration is described
KTM Tokamak operation scenarios software infrastructure
Energy Technology Data Exchange (ETDEWEB)
Pavlov, V.; Baystrukov, K.; Golobkov, YU.; Ovchinnikov, A.; Meaentsev, A.; Merkulov, S.; Lee, A. [National Research Tomsk Polytechnic University, Tomsk (Russian Federation); Tazhibayeva, I.; Shapovalov, G. [National Nuclear Center (NNC), Kurchatov (Kazakhstan)
2014-10-15
One of the largest problems for tokamak devices such as Kazakhstan Tokamak for Material Testing (KTM) is the operation scenarios' development and execution. Operation scenarios may be varied often, so a convenient hardware and software solution is required for scenario management and execution. Dozens of diagnostic and control subsystems with numerous configuration settings may be used in an experiment, so it is required to automate the subsystem configuration process to coordinate changes of the related settings and to prevent errors. Most of the diagnostic and control subsystems software at KTM was unified using an extra software layer, describing the hardware abstraction interface. The experiment sequence was described using a command language. The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.
Runaway-ripple interaction in Tokamaks
International Nuclear Information System (INIS)
Laurent, L.; Rax, J.M.
1989-08-01
Two approaches of the interaction between runaway electrons and the ripple field, in tokamaks, are discussed. The first approach considers the resonance effect as an intense cyclotron heating of the electrons, by the ripple field, in the guiding center frame of the fast particles. In the second approach, an Hamiltonian formalism is used. A criterion for the onset of chaotic behavior and the results are given. A new universal instability of the runaway population in tokamak configuration is found. When combined with cyclotron losses one of its major consequence is to act as an effective slowing down mechanism preventing the free fall acceleration toward the synchrotron limit. This configuration allows the explanation of some experimental results of Tore Supra and Textor
International Nuclear Information System (INIS)
Edelbaum, T.N.; Serben, S.; Var, R.E.
1976-01-01
A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks
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
Distributed digital real-time control system for the TCV tokamak and its applications
Anand, H.; Galperti, C.; Coda, S.; Duval, B.P.; Felici, F.; Blanken, T.; Maljaars, E.; Moret, J.M.; Sauter, O.; Goodman, T.P.; Kim, D.
2017-01-01
A key feature of the new digital plasma control system installed on the TCV (Tokamak à Configuration Variable) tokamak is its possibility to rapidly design, test and deploy real-time algorithms. It accommodates hundreds of diagnostic inputs and actuator outputs, and offers the possibility to design
Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations
International Nuclear Information System (INIS)
Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.
1986-06-01
Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost
Upgrade of the TCV tokamak, first phase: Neutral beam heating system
Energy Technology Data Exchange (ETDEWEB)
Karpushov, Alexander N., E-mail: alexander.karpushov@epfl.ch [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Alberti, Stefano; Chavan, René [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Davydenko, Vladimir I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Duval, Basil P. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Ivanov, Alexander A. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Fasel, Damien; Fasoli, Ambrogio [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Gorbovsky, Aleksander I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Goodman, Timothy [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Kolmogorov, Vyacheslav V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Martin, Yves; Sauter, Olivier [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Sorokin, Aleksey V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); and others
2015-10-15
Highlights: • Widening the parameter range of reactor relevant regimes on the TCV tokamak. • Installation of 1 MW, 30 keV neutral beam, direct ion heating, access to T{sub i}/T{sub e} ≥ 1. • ASTRA simulation of plasma response to NB and EC heating in different regimes. • Specific low divergency neutral beam injector with tunable beam power and energy. - Abstract: Experiments on TCV are designed to complement the work at large integrated tokamak facilities (such as JET) to provide a stepwise approach to extrapolation to ITER and DEMO in areas where medium-size tokamaks can often exploit their experimental capabilities and flexibility. Improving the understanding and control requirements of burning plasmas is a major scientific challenge, requiring access to plasma regimes and configurations with high normalized plasma pressure and a wide range of ion to electron temperature ratios, including T{sub e}/T{sub i} ∼ 1. These conditions will be explored by adding a 1 MW neutral heating beam to TCV's auxiliary for direct ion heating (2015) and increasing the ECH power injected in X-mode at the third harmonic (2 MW in 2015–2016). The manufacturing of the neutral beam injector was launched in 2014.
The ARIES-ST study: Assessment of the spherical tokamak concept as fusion power plants
International Nuclear Information System (INIS)
Najmabadi, F.; Tillack, M.; Miller, R.; Mau, T.K.; Jardin, S.; Stambaugh, R.; Steiner, D.; Waganer, L.
2001-01-01
Recent experimental achievements and theoretical studies have generated substantial interest in the spherical tokamak concept. The ARIES-ST study was undertaken as a national U.S. effort to investigate the potential of the spherical tokamak concept as a fusion power plant and as a vehicle for fusion development. The 1000-MWe ARIES-ST power plant has an aspect ratio of 1.6, a major radius of 3.2 m, a plasma elongation (at 95% flux surface) of 3.4 and triangularity of 0.64. This configuration attains a β of 54% (which is 90% of the maximum theoretical β). While the plasma current is 31 MA, the almost perfect alignment of bootstrap and equilibrium current density profiles results in a current-drive power of only 31 MW. The on-axis toroidal field is 2.1 T and the peak field at the TF coil is 7.6 T, which leads to 288 MW of Joule losses in the normal-conducting TF system. The ARIES-ST study has highlighted many areas where tradeoffs among physics and engineering systems are critical in determining the optimum regime of operation for spherical tokamaks. Many critical issues also have been identified which must be resolved in R and D programs. (author)
Quantitative spectroscopy of X-ray lines and continua in tokamaks
International Nuclear Information System (INIS)
Peacock, N.J.; Lawson, K.D.; Patel, A.; Barnsley, R.; Melnick, I.M.; O'Mullane, M.G.; Singleton, M.A.
1996-10-01
Crystal and synthetic multilayer diffractors, deployed either as flat Bragg reflectors, or curved, as in the Johann configuration, are used to study the spectrum of COMPASS-D and other tokamaks in the wavelength region 1-100 A. In this paper we concentrate on the measurement of absolute photon fluxes and the derivation of volume emissivities of the lines and continua in the X-ray region. The sensitivities of these instruments to absolute photon flux have been constructed ab initio from the individual component efficiencies, including published values of the diffractor reflectivities, which have been checked or supplemented by measurements using a double-axis goniometer or from line branching ratios. For these tokamak plasmas where the elemental abundances and effective ion charge are documented, the X-ray continuum intensity itself has been used as a calibration source to derive absolute instrument sensitivity, in reasonable agreement with the ab initio method. In the COMPASS-D tokamak, changes in the effective ion charge state, Z eff , have been derived for different operating conditions, from the absolute intensity of the continuum at ∼4 A. From the irradiances of the line emission, changes in the absolute level of impurities following ''boronisation'' of the vacuum vessel have also been documented. (UK)
System assessment of helical reactors in comparison with tokamaks
International Nuclear Information System (INIS)
Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.
2002-10-01
A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)
'Crescent'-shaped tokamak for compact ignition
International Nuclear Information System (INIS)
Yamazaki, K.; Reiersen, W.T.
1985-12-01
A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross-section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design dose not rely on the second stability and is easily extendable from the present dee-shaped design. (author)
'Crescent'-shaped tokamak for compact ignition
International Nuclear Information System (INIS)
Yamazaki, K.; Reiersen, W.T.
1986-01-01
A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design does not rely on the second stability and is easily extendable from the present dee-shaped design. (author)
International Nuclear Information System (INIS)
Wesson, John; Campbell, D.J.; Connor, J.W.
1997-01-01
It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)
Recent advances in the HL-2A tokamak experiments
International Nuclear Information System (INIS)
Liu, Y.; Ding, X.T.; Yang, Q.W.; Yan, L.W.; Liu, D.Q.; Xuan, W.M.; Chen, L.Y.; Song, X.M.; Cao, Z.; Zhang, J.H.; Mao, W.C.; Zhou, C.P.; Li, X.D.; Wang, S.J.; Yan, J.C.; Bu, M.N.; Chen, Y.H.; Cui, C.H.; Cui, Z.Y.; Deng, Z.C.; Hong, W.Y.; Hu, H.T.; Huang, Y.; Kang, Z.H.; Li, B.; Li, W.; Li, F.Z.; Li, G.S.; Li, H.J.; Li, Q.; Li, Y.G.; Li, Z.J.; Liu, Yi; Liu, Z.T.; Luo, C.W.; Mao, X.H.; Pan, Y.D.; Rao, J.; Shao, K.; Song, X.Y.; Wang, M.; Wang, M.X.; Wang, Q.M.; Xiao, Z.G.; Xie, Y.F.; Yao, L.H.; Yao, L.Y.; Zheng, Y.J.; Zhong, G.W.; Zhou, Y.; Pan, C.H.
2005-01-01
Two experiment campaigns were conducted on the HL-2A tokamak in 2003 and 2004 after the first plasma was obtained at the end of 2002. Progresses in many aspects have been made, especially in the divertor discharge and feedback control of plasma configuration. Up to now, the following operation parameters have been achieved: I p = 320 kA, B t = 2.2 T and discharge duration T d = 1580 ms. With the feedback control of plasma current and horizontal position, an excellent repeatability of the discharge has been achieved. The tokamak has been operated at both limiter configuration and single null (SN) divertor configuration. The HL-2A SN divertor configuration is simulated with the MHD equilibrium code SWEQU. When the divertor configuration is formed, the impurity radiation in the main plasma decreases remarkably
International Nuclear Information System (INIS)
Baker, C.C.
1981-01-01
This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features
International Nuclear Information System (INIS)
Murray, J.G.; Gorker, G.E.
1985-01-01
Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage
International Nuclear Information System (INIS)
Bickerton, R.J.
1977-01-01
The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)
Energy Technology Data Exchange (ETDEWEB)
Murray, J.G.; Gorker, G.E.
1985-01-01
Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.
Enhancement of confinement in tokamaks
International Nuclear Information System (INIS)
Furth, H.P.
1986-05-01
A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime
International Nuclear Information System (INIS)
Iguchi, H.; Ida, K.; Yamada, H.
1994-01-01
Radial particle transport has been experimentally studied in the low-aspect-ratio heliotron/torsatron device CHS. A non-diffusive outward particle flow (inverse pinch) is observed in the magnetic configuration with the magnetic axis shifted outward, while an inward pinch, like in tokamaks, is observed with the magnetic axis shifted inward. This change in the direction of anomalous particle flow is not due to the reversal of temperature gradient nor the radial electric field. The observation suggests that the particle pinch velocity is sensitive to the magnetic field structure. (author)
HL-1 tokamak data acquisition system and its initial application in the physical experiment
International Nuclear Information System (INIS)
Deng Huichen; Fu Bo; Dong Jiafu
1989-11-01
A HL-1 tokamak data acquisition system has been developed and has been used in the physical experiment. The hardware and software configuration of the system, as well as the typical acquired data in the HL-1 experiment are introduced
Half- coalescence of the m/n = 1 magnetic island in Tokamaks
International Nuclear Information System (INIS)
Bussac, M.N.; Pellat, R.
1986-01-01
We show that a configuration containing an m/n = 1 magnetic island is unstable to an ideal MHD mode. The expected nonlinear implications of this instability could explain the disruptive phase of the classical sawtooth behaviour of Tokamak plasmas
Mechanical configuration and maintenance
International Nuclear Information System (INIS)
Brown, T.G.; Casini, G.; Churakov, G.F.
1982-01-01
The INTOR engineering design has been strongly influenced by considerations for assembly and maintenance. A maintenance philosophy was established at the outset of the conceptual design to insure that the tokamak configuration would be developed to accommodate maintenance requirements. The main features of the INTOR design are summarized in this paper with primary emphasis on the impact of maintenance considerations. The most apparent configuration design feature is the access provided for torus maintenance. Particular attention was given to the size and location of superconducting magnets and the location of vacuum boundaries. All of the poloidal field (PF) coils are placed outside of the bore of the toroidal field (TF) coils and located above and below an access opening between adjacent TF coils through which torus sectors are removed. A magnet structural configuration consisting of mechanically attached reinforcing members has been designed which facilitates the open access space for torus sector removal. For impurity control, a single null poloidal divertor was selected over a double null design in order to maintain sufficient access for pumping and maintenance of the collector. A double null divertor was found to severely limit access to the torus with the addition of divertor collectors and pumping at the top. For this reason, a single null concept was selected in spite of the more difficult design problems associated with the required asymmetric PF system and higher particle loadings
Multi-mode remote participation on the GOLEM tokamak
International Nuclear Information System (INIS)
Svoboda, V.; Huang, B.; Mlynar, J.; Pokol, G.I.; Stoeckel, J.; Vondrasek, G.
2011-01-01
The GOLEM tokamak (formerly CASTOR) at Czech Technical University is demonstrated as an educational tokamak device for domestic and foreign students. Remote participation of several foreign universities (in Hungary, Belgium, Poland and Costa Rica) has been successfully performed. A unique feature of the GOLEM device is functionality which enables complete remote participation and control, solely through Internet access. Basic remote control is possible either in online mode via WWW/SSH interface or offline mode using batch processing code. Discharge parameters are set in each case to configure the tokamak for a plasma discharge. Using the X11 protocol it is possible to control in an advanced mode many technological aspects of the tokamak operation, including: i) vacuum pump initialization, ii) chamber baking, iii) charging of power supplies, iv) plasma discharge scenario, v) data acquisition system.
Kewlani, Gaurav
2016-03-24
Turbulent premixed combustion is studied using experiments and numerical simulations in an acoustically uncoupled cylindrical sudden-expansion swirl combustor, and the impact of the equivalence ratio on the flame–flow characteristics is analyzed. In order to numerically capture the inherent unsteadiness exhibited in the flow, the large eddy simulation (LES) technique based on the artificial flame thickening combustion model is employed. The experimental data are obtained using particle image velocimetry. It is observed that changes in heat loading, in the presence of wall confinement, significantly influence the flow field in the wake region, the stabilization location of the flame, and the flame intensity. Specifically, increasing the equivalence ratio drastically reduces the average inner recirculation zone size and causes transition of the flame macrostructure from the “V” configuration to the “M” configuration. In other words, while the flame stabilizes along the inner shear layer for the V flame, a persistent diffuse reaction zone is also manifested along the outer shear layer for the M flame. The average chemiluminescence intensity increases in the case of the M flame macrostructure, while the axial span of the reaction zone within the combustion chamber decreases. The predictions of the numerical approach resemble the experimental observations, suggesting that the LES framework can be an effective tool for examining the effect of heat loading on flame–flow interactions and the mechanism of transition of the flame macrostructure with a corresponding change in the equivalence ratio.
Tokamak engineering test reactor
International Nuclear Information System (INIS)
Conn, R.W.; Jassby, D.L.
1975-07-01
The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson
1998-12-31
This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson
1997-12-31
This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.
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
HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T
International Nuclear Information System (INIS)
Schultz, J.H.; Montgomery, D.B.
1983-04-01
HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described
The residual zonal dynamics in a toroidally rotating tokamak
International Nuclear Information System (INIS)
Zhou Deng
2015-01-01
Zonal flows, initially driven by ion-temperature-gradient turbulence, may evolve due to the neoclassic polarization in a collisionless tokamak plasma. In this presentation, the form of the residual zonal flow is presented for tokamak plasmas rotating toroidally at arbitrary velocity. The gyro-kinetic equation is analytically solved to give the expression of residual zonal flows with arbitrary rotating velocity. The zonal flow level decreases as the rotating velocity increases. The numerical evaluation is in good agreement with the previous simulation result for high aspect ratio tokamaks. (author)
Compact tokamak reactors. Part 1 (analytic results)
International Nuclear Information System (INIS)
Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.
1996-01-01
We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model
Compact Ignition Tokamak conventional facilities optimization
International Nuclear Information System (INIS)
Commander, J.C.; Spang, N.W.
1987-01-01
A high-field ignition machine with liquid-nitrogen-cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed for the next phase of the United States magnetically confined fusion program. A team of national laboratory, university, and industrial participants completed the conceptual design for the CIT machine, support systems and conventional facilities. Following conceptual design, optimization studies were conducted with the goal of improving machine performance, support systems design, and conventional facilities configuration. This paper deals primarily with the conceptual design configuration of the CIT conventional facilities, the changes that evolved during optimization studies, and the revised changes resulting from functional and operational requirements (F and ORs). The CIT conventional facilities conceptual design is based on two premises: (1) satisfaction of the F and ORs developed in the CIT building and utilities requirements document, and (2) the assumption that the CIT project will be sited at the Princeton Plasma Physics Laboratory (PPPL) in order that maximum utilization can be made of existing Tokamak Fusion Test Reactor (TFTR) buildings and utilities. The optimization studies required reevaluation of the F and ORs and a second look at TFTR buildings and utilities. Some of the high-cost-impact optimization studies are discussed, including the evaluation criteria for a change from the conceptual design baseline configuration. The revised conventional facilities configuration are described and the estimated cost impact is summarized
International Nuclear Information System (INIS)
Morcos, A.; Taylor, H. S.
1989-01-01
This paper will briefly discuss the reason for and content of configuration management both for new plants and, when adapted, for older plants. It will then address three types of activities a utility may undertake as part of a nuclear CAM program and with which Sargent and Leyden has been actively involved. The first activity is a methodology for preparing design-basis documentation. The second is the identification of essential data required to be kept by the utility in support of the operation of a nuclear plant. The third activity is a computerized classification system of plant components, allowing ready identification of plant functional and physical characteristics. Plant configuration documentation describes plant components, the ways they arranged to interact, and the ways they are enabled to interact. Configuration management, on the other hand, is more than the control of such documentation. It is a dynamic process for ensuring that a plant configuration meets all relevant requirements for safety and economy, even while the configuration changes and even while the requirements change. Configuration management for a nuclear plant is so complex that it must be implemented in phases and modules. It takes advantage of and integrates existing programs. Managing complexity and streamlining the change process become important additional objectives of configuration management. The example activities fulfill essential goals of an overall CAM program: definition of design baseline, definition of essential plant data, and classification of plant components
The tokamak - an imperfect frame of refernce
International Nuclear Information System (INIS)
Schmitter, K.U.
1981-03-01
It is attempted to assess the suitability of tokamaks for fusion power plants on the basis of existing design studies by reference to the reality of energy production in fission power plants. A definition of suitability criteria and a discussion of their relation to the most important features of power plants are followed by a comparative treatment. For example, the mean volumetric net electric power density in the nuclear islands of tokamak power plant designs is only 2,5 to 4 E of the value common today in light water reactor nuclear islands. In addition, configuration problems, auxiliary power requirements and energy payback time are discussed and taken into account in the assessment. (orig.)
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
Constrained ripple optimization of Tokamak bundle divertors
International Nuclear Information System (INIS)
Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.
1983-02-01
Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ω B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple ( 0 ) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded
Joint research using small tokamaks
International Nuclear Information System (INIS)
Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van
2005-01-01
Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)
Joint research using small tokamaks
International Nuclear Information System (INIS)
Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.
2005-01-01
Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented
Diagnostics systems for the TBR-E tokamak
International Nuclear Information System (INIS)
Ueda, M.; Ferreira, J.L.; Aso, Y.; Ferreira, J.G.
1992-08-01
A general view of the several diagnostics systems proposed for the TBR-E tokamak is given. This project is a joint undertaking of INPE, USP and UNICAMP plasma laboratories. The requirements for the measurements of the plasma produced parameters are described. Special attention is given for diagnostics used to investigate new physical issues on a low aspect ratio tokamak such as TBR-E. (author)
Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR
Energy Technology Data Exchange (ETDEWEB)
Sa, Wanderley Pires de
1996-12-31
The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.
Neutronic scoping studies for the tokamak experimental power reactor
International Nuclear Information System (INIS)
Santoro, R.T.; Bettis, E.S.; McAlees, D.G.; Watts, H.L.; Williams, M.L.
1976-02-01
One-dimensional neutron and photon radiation transport methods have been used to investigate candidate blanket configurations and compositions for use in the Tokamak Experimental Power Reactor. Seven blanket designs are compared in terms of energy recovery, radiation attenuation, potential radiation damage, and, where applicable, tritium breeding
Fundamental studies on multicharged impurities of the Tokamaks
International Nuclear Information System (INIS)
Bernardi, J. de; Geller, R.; Jacquot, B.; Jacquot, C.
1979-11-01
We describe Micromafios which is an Ion source delivering particles up to completely stripped Carbon, Nitrogen and Oxygen ions. The beams are utilized on an experiment device for cross section measurements enabling fondamental studies for Tokamak impurity ions. The source is a miniaturized Supermafios device where the minimum B configuration is obtained with Samarium Cobalt magnetized Ferrites and a solenoidal axial field [fr
Stirling Engine Configuration Selection
Directory of Open Access Journals (Sweden)
Jose Egas
2018-03-01
Full Text Available Unlike internal combustion engines, Stirling engines can be designed to work with many drive mechanisms based on the three primary configurations, alpha, beta and gamma. Hundreds of different combinations of configuration and mechanical drives have been proposed. Few succeed beyond prototypes. A reason for poor success is the use of inappropriate configuration and drive mechanisms, which leads to low power to weight ratio and reduced economic viability. The large number of options, the lack of an objective comparison method, and the absence of a selection criteria force designers to make random choices. In this article, the pressure—volume diagrams and compression ratios of machines of equal dimensions, using the main (alpha, beta and gamma crank based configurations as well as rhombic drive and Ross yoke mechanisms, are obtained. The existence of a direct relation between the optimum compression ratio and the temperature ratio is derived from the ideal Stirling cycle, and the usability of an empirical low temperature difference compression ratio equation for high temperature difference applications is tested using experimental data. It is shown that each machine has a different compression ratio, making it more or less suitable for a specific application, depending on the temperature difference reachable.
Computational study of axisymmetric modes in noncircular cross section tokamaks
International Nuclear Information System (INIS)
Johnson, J.L.; Chance, M.S.; Greene, J.M.; Grimm, R.C.; Jardin, S.C.; Kerner, W.; Manickam, J.; Weimer, K.E.
1976-09-01
A major computational program to investigate the MHD equilibrium, stability, and nonlinear evolution properties of realistic tokamak configurations is proceeding. Preliminary application is made to the Princeton PDX device. Both axisymmetric (n = 0) modes and kink (n = 1) modes are found; the growth rates depend sensitively on the configuration. A study of the nonlinear evolution of axisymmetric modes in such a device shows that flux conservation in the vacuum region can limit their growth
Steady-state resistive toroidal-field coils for tokamak reactors
International Nuclear Information System (INIS)
Kalnavarns, J.; Jassby, D.L.
1979-12-01
If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m 2 , minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems
Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor
Directory of Open Access Journals (Sweden)
Bong Guen Hong
2018-02-01
Full Text Available A configuration of a fusion-driven transmutation reactor with a low aspect ratio tokamak-type neutron source was determined in a self-consistent manner by using coupled analysis of tokamak systems and neutron transport. We investigated the impact of blanket configuration on the characteristics of a fusion-driven transmutation reactor. It was shown that by merging the TRU burning blanket and tritium breeding blanket, which uses PbLi as the tritium breeding material and as coolant, effective transmutation is possible. The TRU transmutation capability can be improved with a reduced blanket thickness, and fast fluence at the first wall can be reduced. Article History: Received: July 10th 2017; Received: Dec 17th 2017; Accepted: February 2nd 2018; Available online How to Cite This Article: Hong, B.G. (2018 Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor. International Journal of Renewable Energy Development, 7(1, 65-70. https://doi.org/10.14710/ijred.7.1.65-70
Conceptual design of a commercial tokamak reactor using resistive magnets
International Nuclear Information System (INIS)
LeClaire, R.J. Jr.
1988-01-01
The future of the tokamak approach to controlled thermonuclear fusion depends in part on its potential as a commercial electricity-producing device. This potential is continually being evaluated in the fusion community using parametric, system, and conceptual studies of various approaches to improving tokamak reactor design. The potential of tokamaks using resistive magnets as commercial electricity-producing reactors is explored. Parametric studies have been performed to examine the major trade-offs of the system and to identify the most promising configurations for a tokamak using resistive magnets. In addition, a number of engineering issues have been examined including magnet design, blanket/first-wall design, and maintenance. The study indicates that attractive design space does exist and presents a conceptual design for the Resistive Magnet Commercial Tokamak Reactor (RCTR). No issue has been identified, including recirculating power, that would make the overall cost of electricity of RCTR significantly different from that of a comparably sized superconducting tokamak. However, RCTR may have reliability and maintenance advantages over commercial superconducting magnet devices
Total magnetic reconnection during a tokamak major disruption
International Nuclear Information System (INIS)
Goetz, J.A.
1990-09-01
Magnetic reconnection has long been considered to be the cause of sawtooth oscillations and major disruptions in tokamak experiments. Experimental confirmation of reconnection models has been hampered by the difficulty of direct measurement of reconnection, which would involve tracing field lines for many transits around the tokamak. Perhaps the most stringent test of reconnection in a tokamak involves measurement of the safety factor q. Reconnection arising from a single helical disturbance with mode numbers m and n should raise q to m/n everywhere inside of the original resonant surface. Total reconnection should also flatten the temperature and current density profiles inside of this surface. Disruptive instabilities have been studied in the Tokapole 2, a poloidal divertor tokamak. When Tokapole 2 is operated in the material limiter configuration, a major disruption results in current termination as in most tokamaks. However, when operated in the magnetic limiter configuration current termination is suppressed and major disruptions appear as giant sawtooth oscillations. The objective of this thesis is to determine if total reconnection is occurring during major disruptions. To accomplish this goal, the poloidal magnetic field has been directly measured in Tokapole 2 with internal magnetic coils. A full two-dimensional measurement over the central current channel has been done. From these measurements, the poloidal magnetic flux function is obtained and the magnetic surfaces are plotted. The flux-surface-averaged safety factor is obtained by integrating the local magnetic field line pitch over the experimentally obtained magnetic surface
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.
Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation
International Nuclear Information System (INIS)
Riedel, K.S.
1989-03-01
This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately
International Nuclear Information System (INIS)
Anon.
1988-01-01
The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables
International Nuclear Information System (INIS)
Wootton, A.J.
1993-04-01
This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported
Magnetic ''islandography'' in tokamaks
International Nuclear Information System (INIS)
Callen, J.D.; Waddell, B.V.; Hicks, H.R.
1978-09-01
Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks
Accelerator technology in tokamaks
International Nuclear Information System (INIS)
Kustom, R.L.
1977-01-01
This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems
International Nuclear Information System (INIS)
Steiner, D.; Embrechts, M.
1990-07-01
This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein
Internal disruption in tokamaks
International Nuclear Information System (INIS)
Kuvshinov, B.N.; Savrukhin, P.V.
1990-01-01
A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described
International Nuclear Information System (INIS)
Unterberg, Bernhard; Samm, Ulrich
2004-01-01
An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas
Internal disruptions in tokamaks
International Nuclear Information System (INIS)
Kuvshinov, B.N.; Savrukhin, P.V.
1990-01-01
Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described
Accessibility of second regions of stability in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Manickam, J.
1985-12-01
Second regions of stability to the ideal ballooning modes have been shown to exist in large-aspect-ratio circular and small-aspect-ratio bean-shaped tokamaks. We report on the existence of these second stability regions in finite-aspect-ratio dee-shaped tokamaks. We also report on the discovery of a second-stable region with respect to the n = 1 external kink mode in a bean-shaped plasma. The role of the shear and current profile in determining these regions of parameter space are discussed. 13 refs., 6 figs.
Accessibility of second regions of stability in tokamaks
International Nuclear Information System (INIS)
Manickam, J.
1985-12-01
Second regions of stability to the ideal ballooning modes have been shown to exist in large-aspect-ratio circular and small-aspect-ratio bean-shaped tokamaks. We report on the existence of these second stability regions in finite-aspect-ratio dee-shaped tokamaks. We also report on the discovery of a second-stable region with respect to the n = 1 external kink mode in a bean-shaped plasma. The role of the shear and current profile in determining these regions of parameter space are discussed. 13 refs., 6 figs
Iron forbidden lines in tokamak discharges
International Nuclear Information System (INIS)
Suckewer, S.; Hinnov, E.
1979-03-01
Several spectrum lines from forbidden transitions in the ground configurations of highly ionized atoms have been observed in the PLT tokamak discharges. Such lines allow localized observations, in the high-temperature regions of the plasma, of ion-temperatures, plasma motions, and spatial distributions of ions. Measured absolute intensities of the forbidden lines have been compared with simultaneous observations of the ion resonance lines and with model calculations in order to deduce the mechanism of level populaions by means of electron collisions and radiative transitions
Iron forbidden lines in tokamak discharges
Energy Technology Data Exchange (ETDEWEB)
Suckewer, S.; Hinnov, E.
1979-03-01
Several spectrum lines from forbidden transitions in the ground configurations of highly ionized atoms have been observed in the PLT tokamak discharges. Such lines allow localized observations, in the high-temperature regions of the plasma, of ion-temperatures, plasma motions, and spatial distributions of ions. Measured absolute intensities of the forbidden lines have been compared with simultaneous observations of the ion resonance lines and with model calculations in order to deduce the mechanism of level populaions by means of electron collisions and radiative transitions.
'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.
Snowflake Divertor Configuration in NSTX
International Nuclear Information System (INIS)
Soukhanovskii, V.A.; Ahn, Joonwook; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.; Maingi, Rajesh; Maqueda, R.J.; McLean, Adam G.; Menard, J.E.; Mueller, D.; Paul, S.F.; Raman, R.; Roquemore, L.; Ryutov, D.D.; Scott, H.A.
2011-01-01
Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.
'Snowflake' divertor configuration in NSTX
International Nuclear Information System (INIS)
Soukhanovskii, V.A.; Ahn, J.-W.; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J.E.; Mueller, D.M.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Ryutov, D.D.; Scott, H.A.
2011-01-01
Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.
"Snowflake" divertor configuration in NSTX
Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.
2011-08-01
Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.
Computational methods for stellerator configurations
International Nuclear Information System (INIS)
Betancourt, O.
1992-01-01
This project had two main objectives. The first one was to continue to develop computational methods for the study of three dimensional magnetic confinement configurations. The second one was to collaborate and interact with researchers in the field who can use these techniques to study and design fusion experiments. The first objective has been achieved with the development of the spectral code BETAS and the formulation of a new variational approach for the study of magnetic island formation in a self consistent fashion. The code can compute the correct island width corresponding to the saturated island, a result shown by comparing the computed island with the results of unstable tearing modes in Tokamaks and with experimental results in the IMS Stellarator. In addition to studying three dimensional nonlinear effects in Tokamaks configurations, these self consistent computed island equilibria will be used to study transport effects due to magnetic island formation and to nonlinearly bifurcated equilibria. The second objective was achieved through direct collaboration with Steve Hirshman at Oak Ridge, D. Anderson and R. Talmage at Wisconsin as well as through participation in the Sherwood and APS meetings
International Nuclear Information System (INIS)
Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.
1978-01-01
MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby
Sawtooth phenomena in tokamaks
International Nuclear Information System (INIS)
Kuvshinov, B.N.; Savrukhin, P.V.
1989-01-01
A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs
International Nuclear Information System (INIS)
Pare, V.K.
1983-01-01
Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging
International Nuclear Information System (INIS)
1993-01-01
This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs
International Nuclear Information System (INIS)
Wootton, A.J.
1990-04-01
This paper discusses the following work on the text tokamak: data systems; particle confinement; impurity transport; plasma rotation; runaway electrons; electron cyclotron heating; FIR system; transient transport; internal turbulence; edge turbulence; ion temperature; EML experiments; impurity pellet experiments; MHD experiments and analysis; TEXT Upgrade; and Upgrade diagnostics
The ARIES tokamak fusion reactor study
International Nuclear Information System (INIS)
Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.
1989-01-01
The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab
Particle orbits in W VII-X configurations
International Nuclear Information System (INIS)
Wobig, H.
1987-01-01
It is shown that magnetic coordinates are very convenient for describing particle orbits and neoclassical losses in stellarator studies. In the configurations considered (WVII-A and AS, Helias, Heliac, and BSX) plateau losses can be reduced by a factor of 3-4 compared with an equivalent tokamak. It is possible to reduce plateau losses and bootstrap current simultaneously. The bootstrap current can be made negligibly small. The Heliac configuration shows larger plateau losses than the equivalent tokamak. It is possible to localize trapped particles in a region of minimum radial drift velocity. The Er fields strongly reduce localized particle losses
Energy Technology Data Exchange (ETDEWEB)
Stott, Peter [JET, UK (United Kingdom)
1998-10-01
The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)
International Nuclear Information System (INIS)
Stott, Peter
1998-01-01
The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)
Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR
Soto, Christian; Gonzalez, Jeferson; Carvajal, Johan; Ribeiro, Celso
2013-10-01
The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R loan to our laboratory via NI-Costa Rica. The interface with the energy, gas fueling, and security systems are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.
Present status of Tokamak research
International Nuclear Information System (INIS)
Basu, Jayanta
1991-01-01
The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs
Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator
International Nuclear Information System (INIS)
Menard, J.E.; Bromberg, L.; Brown, T.; Burgess, Thomas W.; Dix, D.; Gerrity, T.; Goldston, R.J.; Hawryluk, R.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G.H.; Neumeyer, C.L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.G.; Zarnstorff, M.C.
2011-01-01
A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.
Physics issues for a very-low-aspect-ratio Quasi-Poloidal Stellarator (QPS)
International Nuclear Information System (INIS)
Lyon, J.F.; Berry, L.A.; Hirshman, S.P.
2003-01-01
A quasi-poloidal stellarator with very low plasma aspect ratio (R/a ∼ 2.7, 1/2-1/4 that of existing stellarators) is a new confinement approach that could ultimately lead to a high-beta compact stellarator reactor. The Quasi-Poloidal Stellarator (QPS) experiment is being developed to test key features of this approach. The QPS will study neoclassical and anomalous transport, stability limits at beta up to 2.5%, the configuration dependence of the bootstrap current, and equilibrium robustness. The quasi-poloidal symmetry leads to neoclassical transport that is much smaller than the anomalous transport. The reduced effective field ripple may also produce reduced poloidal viscosity, enhancing the ambipolar E x B poloidal drift and allowing larger poloidal flows for reduction of anomalous transport. A region of second stability exists in the QPS experiment at higher beta. Very-high-beta configurations with a tokamak-like transform profile have also been obtained with a bootstrap current 1/3-1/5 that in an equivalent tokamak. These configurations are stable to low-n ideal MHD kink and vertical instabilities for beta up to 11%. Ballooning-stable configurations are found for beta in the range 2% to 23%. (author)
International Nuclear Information System (INIS)
Beavers, R.R.; Sumiec, K.F.
1989-01-01
Increasing regulatory and industry attention has been focused on properly controlling electrical design changes. These changes can be controlled by using configuration management techniques. Typically, there are ongoing modifications to various process systems or additions due to new requirements at every power plant. Proper control of these changes requires that an organized method be used to ensure that all important parameters of the electrical auxiliary systems are analyzed and that these parameters are evaluated accurately. This process, commonly referred to as configuration management, is becoming more important on both fossil and nuclear plants. Recent NRC- and utility-initiated inspections have identified problems due to incomplete analysis of changes to electrical auxiliary systems at nuclear stations
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
Stability of Tokamaks with respect to slip motions
International Nuclear Information System (INIS)
Rebhan, E.; Salat, A.
1976-06-01
Using the energy principle in Tokamaks we investigate a class of perturbations which, if unstable, cannot be stabilized by the toroidal main field. On the assumptions of usual Tokamak ordering and in the limit of infinite aspect ratio, these perturbations are shown to be minimizing among all axisymmetric perturbations. In the case of finite aspect ratio, a detailed stability analysis is carried out using a constant pressure surface current model with elliptic, triangular or rectangular plasma cross-section. Definite stabilization by toroidal effects and by beta poloidal is demonstrated. (orig.) [de
Measurements of RF-induced sol modifications in Tore Supra tokamak
International Nuclear Information System (INIS)
Kubic, Martin; Gunn, James P.; Colas, Laurent; Heuraux, Stephane; Faudot, Eric
2012-01-01
Since spring 2011, one of the three ion cyclotron resonance heating (ICRH) antennas in the Tore Supra (TS) tokamak is equipped with a new type of Faraday screen (FS). Results from Radio Frequency (RF) simulations of the new Faraday screen suggest the innovative structure with cantilevered bars and 'shark tooth' openings significantly changes the current flow pattern on the front of the antenna which in turn reduces the RF potential and RF electrical field in particular parallel to the magnetic field lines which contributes to generating RF sheaths. Effects of the new FS operation on RF-induced scrape-off layer (SOL) modifications are studied for different plasma and antenna configurations - scans of strap power ratio imbalance, phasing, injected power and SOL density. (authors)
Stability of elongated cross-section tokamaks to axisymmetric even poloidal mode number deformations
International Nuclear Information System (INIS)
Weiner, R.; Jardin, S.C.; Pomphrey, N.
1989-06-01
A recent paper by Nakayama, Sato and Matsuoka suggests that elliptical cross section tokamaks with aspect ratio R/a = 3.2 and with elongation κ = 2.6 are unstable to a splitting (m = 2, n = 0) instability for plasma β > 5%, and that κ /> =/ 4.0 plasmas are unstable to splitting for β /> =/ 1%. We have tried to reproduce these results using the MHD evolution code TSC, but find these configurations to be stable, not even near a stability boundary. Even a κ = 3.7 plasma with β = 23.0% is stable to the splitting mode. However, the addition of pinching coils at the waist will cause the plasma to split if the current in these coils exceeds a critical value I/sub c/ which decreases with increasing β. 8 refs., 11 figs., 1 tab
The calculation of Tritium burnup in Tokamaks
International Nuclear Information System (INIS)
Bittoni, E.; Haegi, M.
1987-01-01
In a deuterium plasma tokamak, the contained fusion-produced tritons are supposed to be decelerated down to thermalization according to classical Coulomb scattering. A fraction of these fast tritons undergoes the DT fusion reaction producing 14.1 MeV neutrons. It is thus possible to get information on the confinement of these fast tritons by comparing the measured and the calculated ratio of the 14.1 MeV to the 2.45 MeV neutron flux. This report describes the calculation of this flux ratio by means of a numerical Monte Carlo-like code
Ignition probabilities for Compact Ignition Tokamak designs
International Nuclear Information System (INIS)
Stotler, D.P.; Goldston, R.J.
1989-09-01
A global power balance code employing Monte Carlo techniques had been developed to study the ''probability of ignition'' and has been applied to several different configurations of the Compact Ignition Tokamak (CIT). Probability distributions for the critical physics parameters in the code were estimated using existing experimental data. This included a statistical evaluation of the uncertainty in extrapolating the energy confinement time. A substantial probability of ignition is predicted for CIT if peaked density profiles can be achieved or if one of the two higher plasma current configurations is employed. In other cases, values of the energy multiplication factor Q of order 10 are generally obtained. The Ignitor-U and ARIES designs are also examined briefly. Comparisons of our empirically based confinement assumptions with two theory-based transport models yield conflicting results. 41 refs., 11 figs
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.
Axisymmetric tokamak scapeoff transport
International Nuclear Information System (INIS)
Singer, C.E.; Langer, W.D.
1982-08-01
We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed
International Nuclear Information System (INIS)
Tendler, M.
1984-06-01
The energy loss from a tokamak plasma due to neutral hydrogen radiation and recycling is of great importance for the energy balance at the periphery. It is shown that the requirement for thermal equilibrium implies a constraint on the maximum attainable edge density. The relation to other density limits is discussed. The average plasma density is shown to be a strong function of the refuelling deposition profile. (author)
International Nuclear Information System (INIS)
Conn, R.W.
1984-05-01
Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described
Modular tokamak magnetic system
International Nuclear Information System (INIS)
Yang, T.F.
1988-01-01
This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil
International Nuclear Information System (INIS)
Conn, R.W.; California Univ., Los Angeles
1984-01-01
Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)
TPX tokamak construction management
International Nuclear Information System (INIS)
Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.
1995-01-01
A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly
Non-inductive current drive and RF heating in SST-1 tokamak
International Nuclear Information System (INIS)
2000-01-01
Steady state superconducting tokamak (SST-1) machine is being developed for 1000 sec operation at different operating parameters. Radio Frequency (RF) and neutral beam injection (NBI) methods are planned in SST-1 for noninductive current drive and heating. In this paper, we describe the non-inductive current drive and RF heating methods that are being developed for this purpose. SST-1 is a large aspect ratio tokamak configured to run double-null divertor plasmas with significant elongation (κ = 1.7-1.9) and triangularity (δ = 0.4-0.7). SST-1 has a major radius of 1.1 in and minor radius of 0.2 m. Circular and shaped plasma experiments would be conducted at 1.5 and 3 T toroidal magnetic field in three different phases with I p = 110 kA and 220 kA. Two main factors have been considered during the development of auxiliary systems, namely, high heat flux (1 MW/m 2 ) incident on the plasma facing antennae components and fast feedback for constant power input due to small energy confinement time (∼ 10 ms). (author)
Stability analysis for the Big Dee upgrade of the Doublet III tokamak
International Nuclear Information System (INIS)
Helton, F.J.; Luxon, J.L.
1987-01-01
Ideal magnetohydrodynamic stability analysis has been carried out for configurations expected in the Big Dee tokamak, an upgrade of the Doublet III tokamak into a non-circular cross-section device which began operation early in 1986. The results of this analysis support theoretical predictions as follows: Since the maximum value of beta stable to ballooning and Mercier modes, which we denote β c , increases with inverse aspect ratio, elongation and triangularity, the Big Dee is particularly suited to obtain high values of β c and there exist high β c Big Dee equilibria for large variations in all relevant plasma parameters. The beta limits for the Big Dee are consistent with established theory as summarized in present scaling laws. High beta Big Dee equilibria are continuously accessible when approached through changes in all relevant input parameters and are structurally stable with respect to variations of input plasma parameters. Big Dee beta limits have a smooth dependence on plasma parameters such as β p and elongation. These calculations indicate that in the actual running of the device the Big Dee high beta equilibria should be smoothly accessible. Theory predicts that the limiting plasma parameters, such as beta, total plasma current and plasma pressure, which can be obtained within the operating limits of the Big Dee are reactor relevant. Thus the Big Dee should be able to use its favourable ideal MHD scaling and controlled plasma shaping to attain reactor relevant parameters in a moderate sized device. (author)
Potential minimum cost of electricity of superconducting coil tokamak power reactors
International Nuclear Information System (INIS)
Reid, R.L.; Peng, Y-K. M.
1989-01-01
The potential minimum cost of electricity (COE) for superconducting tokamak power reactors is estimated by increasing the physics (confinement, beta limit, bootstrap current fraction) and technology [neutral beam energy, toroidal field (TF) coil allowable stresses, divertor heat flux, superconducting coil critical field, critical temperature, and quench temperature rise] constraints far beyond those assumed for ITER until the point of diminishing returns is reached. A version of the TETRA systems code, calibrated with the ITER design and modified for power reactors, is used for this analysis, limiting this study to reactors with the same basic device configuration and costing algorithms as ITER. A minimum COE is reduced from >200 to about 80 mill/kWh when the allowable design constraints are raised to 2 times those of ITER. At 4 times the ITER allowables, a minimum COE of about 60 mill/kWh is obtained. The corresponding tokamak has a major radius of approximately 4 m, a plasma current close to 10 MA, an aspect ratio of 4, a confinement H- factor ≤3, a beta limit of approximately 2 times the first stability regime, a divertor heat flux of about 20 MW/m 2 , a Β max ≤ 18 T, and a TF coil average current density about 3 times that of ITER. The design constraints that bound the minimum COE are the allowable stresses in the TF coil, the neutral beam energy, and the 99% bootstrap current (essentially free current drive). 14 refs., 4 figs., 2 tabs
Configuration studies of LHD plasmas
International Nuclear Information System (INIS)
Okamoto, M.
1997-01-01
Configuration studies are performed on the plasmas of The Large Helical Device (LHD), the construction of which is almost completed at the National Institute for Fusion Science. The LHD has flexibility as an experimental device and can have various configurations by changing the poloidal magnetic fields, the pitch of the helical coils (pitch parameter), and the ratio of currents flowing in the two helical coils. Characteristics of the plasma are investigated for the standard configuration, the change in the pitch parameter, and the helical axis configuration
Configuration studies of LHD plasmas
Energy Technology Data Exchange (ETDEWEB)
Okamoto, Masao
1997-03-01
Configuration studies are performed on the plasmas of The Large Helical Device (LHD), the construction of which is almost completed at the National Institute for Fusion Science. The LHD has flexibility as an experimental device and can have various configurations by changing the poloidal magnetic fields, the pitch of the helical coils (pitch parameter), and the ratio of currents flowing in the two helical coils. Characteristics of the plasma are investigated for the standard configuration, the change in the pitch parameter, and the helical axis configuration. (author)
Study on assembly techniques and procedures for ITER tokamak device
International Nuclear Information System (INIS)
Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi; Ue, Koichi; Shimizu, Katsusuke; Onozuka, Masanori
2006-06-01
The International Thermonuclear Experimental Reactor (ITER) tokamak is mainly composed of a doughnut-shaped vacuum vessel (VV), four types of superconducting coils such as toroidal field coils (TF coils) arranged around the VV, and in-vessel components, such as blanket and divertor. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of the VV and the TF coil are required to be a high accuracy of ±3 mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements as well as the configuration of the tokamak with large size and heavy weight. Based on the above backgrounds, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The tokamak assembly operations are categorized into six work break down structures (WBS), i.e., (1) preparation for assembly operations, (2) sub-assembly of the 40deg sector composed of 40deg VV sector, two TF coils and thermal shield between VV and TF coil at the assembly hall, (3) completion of the doughnut-shaped tokamak assembly composed of nine 40deg sectors in the cryostat at the tokamak pit, (4) measurement of positioning and accuracy after the completion of the tokamak assembly, (5) installation of the ex-vessel components, and (6) installation of in-vessel components. In the present report, two assembly operations of (2) and (3) in the above six WBS, which are the most critical in the tokamak assembly, are mainly described. The report describes the following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology
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
International Nuclear Information System (INIS)
Alladio, F.
1985-01-01
A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs
The evolution of configuration from q > 1 to q < 1
International Nuclear Information System (INIS)
Zhang Peng
1993-06-01
The evolution of configuration from an initial state of tokamak-like plasma to RFP (reversed field pinch) along the trajectory of minimum energy state is studied. the high plasma current allowed in a RFP is expected to be sufficient to heat the plasma to ignition without the need of auxiliary neutral-beam or radio-frequency heating
International Nuclear Information System (INIS)
Rogister, A L
2004-01-01
John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to
The tokamak as a neutron source
International Nuclear Information System (INIS)
Hendel, H.W.; Jassby, D.L.
1989-11-01
This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs
The magnet system of the Tokamak T-15 upgrade
International Nuclear Information System (INIS)
Khvostenko, P.P.; Azizov, E.A.; Alfimov, D.E.; Belyakov, V.A.; Bondarchuk, E.N.; Chudnovsky, A.N.; Dokuka, V.N.; Kavin, A.A.; Khayrutdinov, R.R.; Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N.; Lukash, V.E.; Mineev, A.B.; Muratov, V.P.
2015-01-01
Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.
The magnet system of the Tokamak T-15 upgrade
Energy Technology Data Exchange (ETDEWEB)
Khvostenko, P.P., E-mail: ppkhvost@rambler.ru [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Azizov, E.A.; Alfimov, D.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Belyakov, V.A.; Bondarchuk, E.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Chudnovsky, A.N.; Dokuka, V.N. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Kavin, A.A. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Khayrutdinov, R.R. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Lukash, V.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Mineev, A.B.; Muratov, V.P. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); and others
2015-10-15
Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.
Equilibrium reconstruction in the TCA/Br tokamak
International Nuclear Information System (INIS)
Sa, Wanderley Pires de
1996-01-01
The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author)
Tokamak instrumentation and controls
International Nuclear Information System (INIS)
Becraft, W.R.; Bettis, E.S.; Houlberg, W.A.; Onega, R.J.; Stone, R.S.
1979-02-01
The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine
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
A self-consistent model of an isothermal tokamak
McNamara, Steven; Lilley, Matthew
2014-10-01
Continued progress in liquid lithium coating technologies have made the development of a beam driven tokamak with minimal edge recycling a feasibly possibility. Such devices are characterised by improved confinement due to their inherent stability and the suppression of thermal conduction. Particle and energy confinement become intrinsically linked and the plasma thermal energy content is governed by the injected beam. A self-consistent model of a purely beam fuelled isothermal tokamak is presented, including calculations of the density profile, bulk species temperature ratios and the fusion output. Stability considerations constrain the operating parameters and regions of stable operation are identified and their suitability to potential reactor applications discussed.
Geodesic acoustic modes in noncircular cross section tokamaks
Energy Technology Data Exchange (ETDEWEB)
Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com; Lakhin, V. P. [National Research Center “Kurchatov Institute,” (Russian Federation); Konovaltseva, L. V. [People’s Friendship University of Russia (Russian Federation); Ilgisonis, V. I. [National Research Center “Kurchatov Institute,” (Russian Federation)
2017-03-15
The influence of the shape of the plasma cross section on the continuous spectrum of geodesic acoustic modes (GAMs) in a tokamak is analyzed in the framework of the MHD model. An expression for the frequency of a local GAM for a model noncircular cross section plasma equilibrium is derived. Amendments to the oscillation frequency due to the plasma elongation and triangularity and finite tokamak aspect ratio are calculated. It is shown that the main factor affecting the GAM spectrum is the plasma elongation, resulting in a significant decrease in the mode frequency.
Numerical simulation for HT-6M tokamak electrical transient behaviours
International Nuclear Information System (INIS)
Yu Yuanqi; Liu Baohua; Pan Yuan
1991-02-01
The following main points are concerned: (1) State equations used for dynamic analysis of all electrical parameters of the tokamak are derived. (2) In order to increase plasma volt-seconds and to get plasma current with longer sustainment phase, a power supply scheme for HT-6M and its numerical simulation are studied. (3) The distribution of energy flow in coupling loops of the tokamak is discussed, and the energy transfer ratio from the OH loop and vertical field loop to the plasma is also analyzed
International Nuclear Information System (INIS)
Komoshvili, K.; Bruma, C.; Cuperman, S.
2004-01-01
Full Text:In the magnetically confined fusion devices, externally launched e.m. waves are used, e.g., for heating, non-inductive current drive and turbulent transport suppression barriers. In view of the complexity of these processes, it is desirable to assist the planning of the actual experiments by reliable theoretical (computational) studies. This work aims to (i) assess the effect of antenna position and extension on the fast waves-plasma interactions in pre-heated spherical tokamaks and consequently, (ii) to further the physical understanding as well as to determine optimal conditions in order to achieve the imposed goals. Thus, using as a study case the spherical tokamak START, we considered the following antenna positions and extensions: (a) low field side location and i T ±π/4 poloidal extension; (b) above and below middle-plane locations (two separate sections) and extending (each) π/2; (c) (hypothetical) circular, 2π-extension. We solved the full wave equations in order to consistently determine the global e.m. field for Alfvinic modes in inhomogeneous, non-uniformly magnetized, resistive, small aspect ratio tokamak plasma in the presence of externally launched fast waves. The global approach consists of simultaneous treatment of the plasma-vacuum-external RF source-vacuum-metal wall configuration with the appropriate consideration of wave propagation, transmission, absorption and mode conversion; in this, no simplifying approximations or small parameter extension are used. Illustrative results of these investigations will be presented and discussed
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
Rotating shield ceiling for the compact ignition tokamak test cell
International Nuclear Information System (INIS)
Commander, J.C.
1986-01-01
For the next phase of the United States fusion program, a compact, high-field, toroidal ignition machine with liquid nitrogen cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed. The CIT machine will be housed in a test cell with design features developed during preconceptual design. Configured as a right cylinder, the selected test cell design features: a test cell and basement with thick concrete shielding walls, and floor; leak tight tritium seals; and operational characteristics well suited to the circular CIT machine configuration and radially oriented ancillary equipment and systems
International Nuclear Information System (INIS)
Rome, J.A.; Peng, Y.K.M.
1978-09-01
Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well
International Nuclear Information System (INIS)
Doggett, J.; Salpietro, E.; Shatalov, G.
1991-01-01
The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs
International Nuclear Information System (INIS)
Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M
2008-01-01
Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 (micro)m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics
Simulation of saturated tearing modes in tokamaks
International Nuclear Information System (INIS)
Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.
2004-01-01
A quasi-linear model, which includes the effect of the neoclassical bootstrap current, is developed for saturated tearing modes in order to compute magnetic island widths in axisymmetric toroidal plasmas with arbitrary aspect ratio and cross-sectional shape. The model is tested in a simple stand-alone code and is implemented in the BALDUR [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1982)] predictive modeling code. It is found that the widths of tearing mode islands increase with decreasing aspect ratio and with increasing elongation. Also, the island widths increase when the gradient of the current density increases at the edge of the islands and when the current density inside the islands is suppressed, such as the suppression caused by the near absence of the bootstrap current within the islands. In simulations of tokamak discharges, it is found that tearing mode island widths oscillate in time in response to periodic sawtooth crashes. The local enhancements in the transport produced by magnetic islands have a noticeable effect on global plasma confinement in simulations of low aspect ratio, high beta tokamaks, where saturated tearing mode islands can occur with widths that are greater than 15% of the plasma minor radius
Modelling dust transport in tokamaks
International Nuclear Information System (INIS)
Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.
2008-01-01
The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)
The stability margin on EAST tokamak
International Nuclear Information System (INIS)
Jin-Ping, Qian; Bao-Nian, Wan; Biao, Shen; Bing-Jia, Xiao; Walker, M.L.; Humphreys, D.A.
2009-01-01
The experimental advanced superconducting tokamak (EAST) is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma due to the necessary thermal isolation from the superconducting magnets, which leads to relatively weaker coupling between plasma and poloidal field. This may cause more difficulties in controlling the vertical instability by using the poloidal coils. The measured growth rates of vertical stability are compared with theoretical calculations, based on a rigid plasma model. Poloidal beta and internal inductance are varied to investigate their effects on the stability margin by changing the values of parameters α n and γ n (Howl et al 1992 Phys. Fluids B 4 1724), with plasma shape fixed to be a configuration with k = 1.9 and δ = 0.5. A number of ways of studying the stability margin are investigated. Among them, changing the values of parameters κ and l i is shown to be the most effective way to increase the stability margin. Finally, a guideline of stability margin M s (κ, l i , A) to a new discharge scenario showing whether plasmas can be stabilized is also presented in this paper
Alfven wave studies on a tokamak
International Nuclear Information System (INIS)
Kortbawi, D.
1987-10-01
The continuum modes of the shear Alfven resonance are studied on the Tokapole II device, a small tokamak operated in a four node poloidal divertor configuration. A variety of antenna designs and the efficiency with which they deliver energy to the resonant layer are discussed. The spatial structure of the driven waves is studied by means of magnetic probes inserted into the current channel. In an attempt to optimize the coupling of energy in to the resonant layer, the angle of antenna currents with respect to the equilibrium field, antenna size, and plasma-to-antenna distance are varied. The usefulness of Faraday shields, particle shields, and local limiters are investigated. Antennas should be well shielded, either a dense Faraday shield or particle shield being satisfactory. The antenna should be large and very near to the plasma. The wave magnetic fields measured show a spatial resonance, the position of which varies with the value of the equilibrium field and mass density. They are polarized perpendicular to the equilibrium field. A wave propagates radially in to the resonant surface where it is converted to the shear Alfven wave. The signal has a short risetime and does not propagate far toroidally. These points are all consistent with a strongly damped shear Alfven wave. Comparisons of this work to theoretical predictions and results from other tokamaks are made
Theoretical scaling law for ohmically heated tokamaks
International Nuclear Information System (INIS)
Minardi, E.
1981-06-01
The electrostatic drift instability arising from the reduction of shear damping, due to toroidal effects, is assumed to be the basic source of the anomalous electron transport in tokamaks. The Maxwellian population of electrons constitutes a medium whose adiabatic nonlinear reaction to the instability (described in terms of an effective dielectric constant of the medium) determines the stationary electrostatic fluctuation level in marginally unstable situations. The existence of a random electrostatic potenial implies a fluctuating current of the Maxwellian electrons which creates a random magnetic field and a stocasticization of a magnetic configuration. The application of recent results allows the calculation of the realted radial electron transport. It is found that the confinement time under stationary ohmic conditions scales as n Tsub(i)sup( - 1/2) and is proportional roughly to the cube of the geometric dimenisions. Moreover, it is deduced that the loop voltage is approximateley the same for all tokamaks, irrespective of temperature and density and to a large extent, also of geometrical conditions. Thes results are characteristic of the ohmic stationary regime and can hardly be extrapolated to order heating regimes. (orig.)
Experimental investigations at the Soviet tokamaks
International Nuclear Information System (INIS)
Bobrovskij, G.A.; Golant, V.E.; AN SSSR, Leningrad. Fiziko-Tekhnicheskij Inst.)
1978-01-01
The review is devoted to the basic results obtained on the Soviet tokamaks during 1976-1977. Behaviour of impurities, tearing instability, additional methods of plasma heating, energy distribution function were investigated. A brief description of new T-7, TM-4, ''Tuman-3'' tokamaks is given. It is shown that despite inflow of impurities to the pinch periphery, no their appreciable accumulation is observed at least during the discharge time. It is shown that the helical perturbations with m=2 and 1 present the greatest danger. The suppression of the tearing instability is related with suppression of the mode with m=2. The helical perturbation prevents formation of skin configuration at the initial stage of the discharge. As a rule, the transition of an appreciable fraction of electrons to continuous acceleration does not take place, although a significant deformation of electron distribution function under the action of electric field occurs. Plasma compression by increasing magnetic field induces oscillations and improves thermal plasma isolation. It is shown experimentally that the considerable efficiency of energy contribution to the ion component at the central part of plasma may be obtained by means of HF heating under conditions of low-hybrid resonance. It is shown that the recombination has a considerable effect on concentration of neutral particles in the central region
MHD stability analysis of axisymmetric surface current model tokamaks close to the spheromak regime
International Nuclear Information System (INIS)
Honma, Toshihisa; Kaji, Ikuo; Fukai, Ichiro; Kito, Masafumi.
1984-01-01
In the toroidal coordinates, a stability analysis is presented for very low-aspect-ratio tokamaks with circular cross section which is described by a surface current model (SCM) of axisymmetric equilibria. The energy principle determining the stability of plasma is treated without any expansion of aspect ratio. Numerical results show that, owing to the occurrence of the non-axisymmetric (n=1) unstable modes, there exists no MHD-stable ideal SCM spheromak characterized by zero external toroidal vacuum field. Instead, a stable spheromak-type plasma which comes to the ideal SCM spheromak is provided by the configuration with a very weak external toroidal field. Close to the spheromak regime (1.0 1 aspect ratio< = 1.1), the minimum safety factor and the critical β-values increase mo notonically with aspect ratio decreasing from a large value, and curves of βsub(p) versus β in the marginal stability approach to an ideal SCM spheromak line βsub(p)=β. (author)
Deposit of thin films for Tokamaks conditioning
International Nuclear Information System (INIS)
Valencia A, R.
2006-01-01
discharge plasma, created in a calibrated mixture of methane-hydrogen during the hydrogenated amorphous carbon film deposit on the vessel wall of Novillo tokamak, were determined by mass spectrometry. By way of measuring the emission lines of the carbon and oxygen impurities in intense discharges, the time required by the plasma to interact with the wall was estimated. In addition to it, the temporal conduct of the emission line intensity of these impurities was observed by means of an intensified CCD detector. Once an ∼ 10 % of helium was introduced in the operating gas of the tokamak discharges, a 25-42 eV time variation of the electron temperature was measured using the intensity ratio technique. (Author)
Controlled thermonuclear fusion and the latest progress on China's HT-7 superconducting tokamak
International Nuclear Information System (INIS)
Li Jiangang; Yang Yu
2003-01-01
After 50 years of research on controlled thermonuclear fusion, a new stage will be reached in 2003, when a site for the International Thermonuclear Experimental Reactor project will be chosen to start the construction. Scientists hope that this project could herald a new era in which the energy problem will be solved completely. The great progress made on the HT-7 superconducting tokamak in China has provided positive and powerful support for fusion research. The HT-7 is one of the only two superconducting tokamaks in the world that can carry out minute-scale high temperature plasma research, and has achieved a duration of 63.95s for the hot plasma discharge. This is a major step towards real steady-state operation of the tokamak configuration. We present an overview of the latest progress on the tokamak experiments in the Institute of Plasma Physics, Chinese Academy of Sciences
Lessons learned from the Tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)
International Nuclear Information System (INIS)
Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Werley, K.A.
1994-01-01
Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety and health (ES ampersand H) characteristics of projected tokamak power plants. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances relative to present understanding in physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advanced tokamak plasmas configured in the second-stability regime that achieve both high β and bootstrap fractions near unity through strong profile control offer high promise in this regard
Tokamak building-design considerations for a large tokamak device
International Nuclear Information System (INIS)
Barrett, R.J.; Thomson, S.L.
1981-01-01
Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release
Conceptual Design of Alborz Tokamak Poloidal Coils System
Mardani, M.; Amrollahi, R.
2013-04-01
The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.
Equilibrium vertical field in the TBR Tokamak
International Nuclear Information System (INIS)
Ueta, A.Y.
1985-01-01
An experimental study on the influence of the vertical magnetic field of the TBR tokamak on the stability and equilibrium of plasma column, was done. Magnetic pick-up coils were built to measure plasma current and position, together with active networks, necessary fo the electronic processing of signals. Some measurements were on the space configuration of the vertical field, and on the influence due to the toroidal vessel. From the data obtained it was possible to discuss the influence of the currents induced on the vessel surface, on plasma equilibrium. Theoretical and experimental results of the vertica field, as a function of plasma current were compared, and allowed an evaluation of the plasma kinetic pressure and temperature. (Author) [pt
Alfven wave coupling in large tokamaks
International Nuclear Information System (INIS)
Borg, G.G.; Knight, A.J.; Lister, J.B.; Appert, K.; Vaclavik, J.
1988-01-01
Supplementary plasma heating by Alfven waves (AWH) has been extensively studied both theoretically and experimentally for small, low temperature plasmas. However, only a few studies of AWH have been performed for fusion plasmas. In this paper the cylindrical kinetic code ISMENE is used to address problems af AWH in a large tokamak. The results of calculations are presented which show that the antenna loading scales with frequency and vessel dimensions according to ideal MHD theory. A sample scaling of the experimental antenna loading measured in TCA to the loading predicted for a fusion plasma is presented. We discuss whether this loading leads to a realistic antenna design. The choice of a suitable antenna configuration, mode number and operating frequency is presented for NET parameters with a typical operating scenario. (author) 6 figs., 8 refs
Natural current profiles in tokamaks
International Nuclear Information System (INIS)
Biskamp, D.
1986-01-01
It is proposed that a certain class of equilibrium, which follow from an elementary variational principle, are the natural current profiles in tokamaks, to which actual discharge profiles tend to relax. (orig.)
Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...
JUST: Joint Upgraded Spherical Tokamak
International Nuclear Information System (INIS)
Azizov, E.A.; Dvorkin, N.Ya.; Filatov, O.G.
1997-01-01
The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed. (author)
Preliminary Design of Alborz Tokamak
Mardani, M.; Amrollahi, R.; Saramad, S.
2012-04-01
The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.
International Nuclear Information System (INIS)
Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.
1991-01-01
A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs
Numerical Tokamak Project code comparison
International Nuclear Information System (INIS)
Waltz, R.E.; Cohen, B.I.; Beer, M.A.
1994-01-01
The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes
Spherical tokamak development in Brazil
International Nuclear Information System (INIS)
Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group
2003-01-01
This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Confinement and diffusion in tokamaks
International Nuclear Information System (INIS)
McWilliams, R.
1988-01-01
The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation
Enhancement of confinement in tokamaks
International Nuclear Information System (INIS)
Furth, H.P.
1986-01-01
The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)
Spherical tokamak development in Brazil
Energy Technology Data Exchange (ETDEWEB)
Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)
2003-07-01
The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Spherical tokamak development in Brazil
Energy Technology Data Exchange (ETDEWEB)
Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group
2003-12-01
This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Spherical tokamak development in Brazil
International Nuclear Information System (INIS)
Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes
2003-01-01
The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
International Nuclear Information System (INIS)
Navratil, G.A.; Mauel, M.E.; Ivers, T.H.; Sankar, M.K.V.; Eisner, E.; Gates, D.; Garofalo, A.; Kombargi, R.; Maurer, D.; Nadle, D.; Xiao, Q.
1993-01-01
During the past 6 months, experiments have been conducted with the HBT-EP tokamak in order to (1) test and evaluate diagnostic systems, (2) establish basic machine operation, (3) document MHD behavior as a function of global discharge parameters, (4) investigate conditions leading to passive stabilization of MHD instabilities, and (5) quantify the external saddle coil current required for DC mode locking. In addition, the development and installation of new hardware systems has occurred. A prototype saddle coil was installed and tested. A five-position (n,m) = (1,2) external helical saddle coil was attached for mode-locking experiments. And, fabrication of the 32-channel UV tomography and the multipass Thomson scattering diagnostics have begun in preparation for installation later this year
Anomalous transport in tokamaks
International Nuclear Information System (INIS)
Wootton, A.J.
1989-01-01
A review is presented of what is known about anomalous transport in tokamaks. It is generally thought that this anomalous transport is the result of fluctuations in various plasma parameters. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation driven fluxes are available. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge energy flux are less clear, but electrostatic convection plays an important part. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made. Magnetic fluctuations is discussed
International Nuclear Information System (INIS)
Tenney, F.H.
1976-09-01
A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters
International Nuclear Information System (INIS)
Tenney, F.H.
1976-01-01
A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters
International Nuclear Information System (INIS)
Schmidt, J.
1987-01-01
The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies
International Nuclear Information System (INIS)
Bondeson, A.
1987-01-01
This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)
The ARIES tokamak reactor study
International Nuclear Information System (INIS)
1989-10-01
The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions
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.
Reducing Turbulent Transport in Toroidal Configurations via Shaping
International Nuclear Information System (INIS)
Mynick, H.E.; Pomphrey, N.; Xanthopoulos, P.
2011-01-01
Recent progress in reducing turbulent transport in stellarators and tokamaks by 3D shaping using a stellarator optimization code in conjunction with a gyrokinetic code is presented. The original applications of the method focussed on ion temperature gradient transport in a quasi-axisymmetric stellarator design. Here, an examination of both other turbulence channels and other starting configurations is initiated. It is found that the designs evolved for transport from ion temperature gradient turbulence also display reduced transport from other transport channels whose modes are also stabilized by improved curvature, such as electron temperature gradient and ballooning modes. The optimizer is also applied to evolving from a tokamak, finding appreciable turbulence reduction for these devices as well. From these studies, improved understanding is obtained of why the deformations found by the optimizer are beneficial, and these deformations are related to earlier theoretical work in both stellarators and tokamaks.
Vertical and horizontal access configurations
International Nuclear Information System (INIS)
Spampinato, P.T.
1987-01-01
A number of configuration features and maintenance operations are influenced by the choice of whether a design is based on vertical or horizontal access for replacing reactor components. The features which are impacted most include the first wall/blanket segmentation, the poloidal field coil locations, the toroidal field coil number and size, access port size for in-vessel components, and facilities. Since either configuration can be made to work, the choice between the two is not clear cut because both have certain advantages. It is apparent that there are large cost benefits in the poloidal field coil system for ideal coil locations for high elongation plasmas and marginal savings for the INTOR case. If we assume that a new tokamak design will require a higher plasma elongation, the recommendation is to arrange the poloidal field coils in a cost-effective manner while providing reasonable midplane access for heating interfaces and test modules. If a new design study is not based on a high elongation plasma, it still appears prudent to consider this approach so that in-vessel maintenance can be accomplished without moving very massive structures such as the bulk shield. 10 refs., 29 figs., 3 tabs
Overview of the ITER Tokamak complex building and integration of plant systems toward construction
Energy Technology Data Exchange (ETDEWEB)
Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others
2015-10-15
The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed
International Nuclear Information System (INIS)
Sinman, S.; Sinman, A.
2003-01-01
The aim of this study is to identify the physical basis of the alternative self-organization mechanism that exists on the STPC-EX machine and to determine complementary features with respect to present compact toroid concepts. In the STPC-EX machine, there exist two solenoids placed inside the central passive floating conductive hollow rod and externally onto flux conserver. These are in a passive state and they do not have an important role in the self-created spherical tokamak plasma (SCSTP) in the STPC-EX machine. In this study, conventional diagnostic tools are used and for photographic recording, the method of open shutter integrated post-fogging is chosen. Two annular coaxial plasma current sheets, one within the other in the same direction, are created and flow on the surface of the central conductive hollow rod. Consequently, the spherical tokamak is configured by a new creation mechanism called the dual-axial z-pinch. High betas of 0.4-0.6 and aspect ratios of up to 1.25 can be obtained. (author)
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
Maintenance concept development for the Compact Ignition Tokamak
International Nuclear Information System (INIS)
Macdonald, D.
1988-01-01
The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs
Design features of HTMR-Hybrid Toroidal Magnet Tokamak Reactor
International Nuclear Information System (INIS)
Rosatelli, F.; Avanzini, P.G.; Brunelli, B.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Zampaglione, V.
1985-01-01
The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfill the scientific and technological objectives expected from next generation devices (e.g. INTOR-NET) with size and costs as small as possible. An hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. In this paper the authors describe the optimization procedure for the hybrid magnet configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils
Design features of HTMR-hybrid toroidal magnet tokamak reactor
International Nuclear Information System (INIS)
Rosatelli, F.; Avanzini, P.G.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Brunelli, B.; Zampaglione, V.
1984-01-01
The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfil the scientific and technological objectives expected from next generation devices with size and costs as small as possible. A hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. The optimization procedure for the hybrid magnet, configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils are described. (author)
Pulse length assessment of compact ignition tokamak designs
International Nuclear Information System (INIS)
Stotler, D.P.; Pomphrey, N.
1989-07-01
A time-dependent zero-dimensional code has been developed to assess the pulse length and auxiliary heating requirements of Compact Ignition Tokamak (CIT) designs. By taking a global approach to the calculation, parametric studies can be easily performed. The accuracy of the procedure is tested by comparing with the Tokamak Simulation Code which uses theory-based thermal diffusivities. A series of runs is carried out at various levels of energy confinement for each of three possible CIT configurations. It is found that for cases of interest, ignition or an energy multiplication factor Q /approxreverse arrowgt/ 7 can be attained within the first half of the planned five-second flattop with 10--40 MW of auxiliary heating. These results are supported by analytic calculations. 18 refs., 7 figs., 2 tabs
Saturated tearing modes in tokamaks with divertors
International Nuclear Information System (INIS)
Bateman, G.
1982-12-01
We have developed a self-consistent theory of saturated tearing modes capable of predicting multiple magnetic island widths in tokamaks with no assumptions on the cross-sectional shape, aspect ratio, or plasma pressure. We are in the process of implementing this algorithm in the form of a computer code. We propose: (1) to complete, refine, document and publish this computer code; (2) to carry out a survey in which we vary the current profile, aspect ratio, cross-sectional shape, and pressure profile in order to determine their effect on saturated tearing mode magnetic island widths; and (3) to determine the effect of some externally applied magnetic perturbation harmonics on these magnetic island widths. Particular attention will be paid to the coupling between different helical harmonics, the effect of multiple magnetic islands on the profiles of temperature, pressure and current, and the potential of magnetic island overlap leading to a disruptive instability
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
Digital controlled pulsed electric system of the ETE tokamak. First report
International Nuclear Information System (INIS)
Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson
1997-01-01
This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author)
Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR
Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso
2013-10-01
The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.
Confinement and exhaust in the Mega Ampere Spherical Tokamak
International Nuclear Information System (INIS)
Counsell, G F; Ahn, J-W; Akers, R; Arends, E; Buttery, R; Field, A R; Gryaznevich, M; Helander, P; Kirk, A; Meyer, H; Valovic, M; Wilson, H R; Yang, Y
2002-01-01
The Mega Ampere Spherical Tokamak (MAST) is now accessing regimes with high normalized confinement relative to international scalings, H H (IPB98(y, 2))>1 at high normalized density, n-bar e >60% of the Greenwald density. Data from MAST H-modes suggest that the aspect ratio dependency of international confinement and L-H threshold scalings may need to be modified to improve predictions for ITER. Access to H-mode on MAST is strongly affected by both the divertor magnetic geometry and fuelling location, with the formation of an edge transport barrier being facilitated by operation near the symmetric, connected double-null configuration and with poloidally localized inboard gas puffing. The ELMs on MAST appear to be Type III in nature, even in the highest performance plasmas and with the maximum available auxiliary heating power. ELM energy losses are less than 4% of stored energy in all regimes so far explored. These Type III ELMs are associated with a reduction in the pedestal density but no significant change in the pedestal temperature or temperature profile, indicating that energy is convected from the pedestal region into the scrape-off layer. Analysis of the energy observed to arrive at the divertor targets indicates that ELM losses are predominantly on the low field side. ELM effluxes are observed up to 20 cm from the plasma edge at the outboard mid-plane and are associated with the radial motion of a feature at an average velocity of 1.2 km s -1
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.
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
Modelling of electron transport and of sawtooth activity in tokamaks
International Nuclear Information System (INIS)
Angioni, C.
2001-10-01
Transport phenomena in tokamak plasmas strongly limit the particle and energy confinement and represent a crucial obstacle to controlled thermonuclear fusion. Within the vast framework of transport studies, three topics have been tackled in the present thesis: first, the computation of neoclassical transport coefficients for general axisymmetric equilibria and arbitrary collisionality regime; second, the analysis of the electron temperature behaviour and transport modelling of plasma discharges in the Tokamak a configuration Variable (TCV); third, the modelling and simulation of the sawtooth activity with different plasma heating conditions. The work dedicated to neoclassical theory has been undertaken in order to first analytically identify a set of equations suited for implementation in existing Fokker-Planck codes. Modifications of these codes enabled us to compute the neoclassical transport coefficients considering different realistic magnetic equilibrium configurations and covering a large range of variation of three key parameters: aspect ratio, collisionality, and effective charge number. A comparison of the numerical results with an analytical limit has permitted the identification of two expressions for the trapped particle fraction, capable of encapsulating the geometrical effects and thus enabling each transport coefficient to be fitted with a single analytical function. This has allowed us to provide simple analytical formulae for all the neoclassical transport coefficients valid for arbitrary aspect ratio and collisionality in general realistic geometry. This work is particularly useful for a correct evaluation of the neoclassical contribution in tokamak scenarios with large bootstrap cur- rent fraction, or improved confinement regimes with low anomalous transport and for the determination of the plasma current density profile, since the plasma conductivity is usually assumed neoclassical. These results have been included in the plasma transport code
Resistive instabilities in tokamaks
International Nuclear Information System (INIS)
Rutherford, P.H.
1985-10-01
Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed
Tokamak experimental power reactor
International Nuclear Information System (INIS)
Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.
1978-01-01
A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars
Classical tokamak transport theory
International Nuclear Information System (INIS)
Nocentini, Aldo
1982-01-01
A qualitative treatment of the classical transport theory of a magnetically confined, toroidal, axisymmetric, two-species plasma is presented. The 'weakly collisional' ('banana' and 'plateau') and 'collision dominated' ('Pfirsch-Schlueter' and 'highly collisional') regimes, as well as the Ware effect are discussed. The method used to evaluate the diffusion coffieicnts of particles and heat in the weakly collisional regime is based on stochastic argument, that requires an analysis of the characteristic collision frequencies and lengths for particles moving in a tokamak-like magnetic field. The same method is used to evaluate the Ware effect. In the collision dominated regime on the other hand, the particle and heat fluxes across the magnetic field lines are dominated by macroscopic effects so that, although it is possible to present them as diffusion (in fact, the fluxes turn out to be proportional to the density and temperature gradients), a macroscopic treatment is more appropriate. Hence, fluid equations are used to inveatigate the collision dominated regime, to which particular attention is devoted, having been shown relatively recently that it is more complicated than the usual Pfirsch-Schlueter regime. The whole analysis presented here is qualitative, aiming to point out the relevant physical mechanisms involved in the various regimes more than to develop a rigorous mathematical derivation of the diffusion coefficients, for which appropriate references are given. (author)
Tokamak experimental power reactor
International Nuclear Information System (INIS)
Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.
1976-01-01
A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors
Optimization study of normal conductor tokamak for commercial neutron source
Fujita, T.; Sakai, R.; Okamoto, A.
2017-05-01
The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.
Dynamic stabilization of disruption precursors in tokamak
Energy Technology Data Exchange (ETDEWEB)
Maoquan, Wang; Jianshan, Mao; Yuan, Pan [Academia Sinica, Hefei, AH (China). Inst. of Plasma Physics
1994-12-01
A method for dynamic stabilization of the disruption precursors in tokamak is proposed, that is a controlled ac current induced and added to the equilibrium current. The ac currents applied can be a sine alternative current with a relevant frequency, or a pulsed current with a suitable pulsed width {tau} and or a discontinuous pulsed current whose width {tau} is very shorter than the intervals between pulses, and or a `sawtooth` pulsed current with the time of ramp phase of the sawtooth is very much shorter than the sawtooth descending time, the ratio of them can be {<=}10{sup -3}. The physical model of the ac current drive is analyzed in detail. The suppression role of the ac current on the MHD perturbations was analyzed in theory and proved numerically. It is indicated that the ac current can make the discontinuous derivative, {Delta}`, more favorable for the tearing mode stabilities, and so, as long as the parameters of the applied ac currents are selected suitably, the MHD perturbations can be suppressed effectively, the perturbations will be in the zero-growing state, the profile of the plasma current and temperature remain in the initial states and not variate basically, the tokamak be in the stabilized operation state. (8 figs.).
Bootstrap currents in stellarators and tokamaks
International Nuclear Information System (INIS)
Okamoto, Masao; Nakajima, Noriyoshi.
1990-09-01
The remarkable feature of the bootstrap current in stellarators is it's strong dependence on the magnetic field configuration. Neoclassical bootstrap currents in a large helical device of torsatron/heliotron type (L = 2, M = 10, R = 4 m, B = 4 T) is evaluated in the banana (1/ν) and the plateau regime. Various vacuum magnetic field configurations are studied with a view to minimizing the bootstrap current. It is found that in the banana regime, shifting of the magnetic axis and shaping of magnetic surfaces have a remarkable influence on the bootstrap current; a small outward shift of the magnetic axis and vertically elongated magnetic surfaces are favourable for a reduction of the bootstrap current. It is noted, however, that the ripple diffusion in the 1/ν regime has opposite tendency to the bootstrap current; it increases with the outward shift and increases as the plasma cross section is vertically elongated. The comparison will be made between bootstrap currents in stellarators and tokamaks. (author)
TCV divertor upgrade for alternative magnetic configurations
Directory of Open Access Journals (Sweden)
H. Reimerdes
2017-08-01
Full Text Available The Swiss Plasma Center (SPC is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.
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)
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)
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)
Core fuelling to produce peaked density profiles in large tokamaks
International Nuclear Information System (INIS)
Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.
1995-01-01
Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. It is shown that with radially 'hollow' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles that are peaked off-axis. The fuelling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fuelling does not require Megavolt particle energies. Even with beam voltages of ∼ 200 keV, however, exceptionally good particle confinement is needed to achieve net electrical power generation. The required ratio of particle to thermal diffusivities is an order of magnitude outside the range reported for tokamaks. In a system with no power production requirement (e.g., neutron sources) neutral beam fuelling should be capable of producing peaked density profiles in devices as large as ITER. Fuelling systems with low energy cost per particle - such as cryogenic pellet injection - must be used in power producing tokamaks when τ P ∼ τ E . Simulations with pellet injection speeds of 7 km/s show that the peaking factor, n e0 / e >, approaches 2. (author). 65 refs, 8 figs
Summary discussion: An integrated advanced tokamak reactor
International Nuclear Information System (INIS)
Sauthoff, N.R.
1994-01-01
The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept
Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance
International Nuclear Information System (INIS)
Nishio, S.; Ueda, S.; Kurihara, R.; Kuroda, T.; Miura, H.; Sako, K.; Takase, H.; Seki, Y.; Adachi, J.; Yamazaki, S.; Hashimoto, T.; Mori, S.; Shinya, K.; Murakami, Y.; Senda, I.; Okano, K.; Asaoka, Y.; Yoshida, T.
2000-01-01
If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal
STARFIRE: a commercial tokamak reactor
International Nuclear Information System (INIS)
1979-12-01
The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor
LHCD experiments on tokamak CASTOR
International Nuclear Information System (INIS)
Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.
1990-10-01
A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs
Reactor Configuration Development for ARIES-CS
International Nuclear Information System (INIS)
Ku LP
2005-01-01
New compact, quasi-axially symmetric stellarator configurations have been developed as part of the ARIES-CS reactor studies. These new configurations have good plasma confinement and transport properties, including low losses of α particles and good integrity of flux surfaces at high β. We summarize the recent progress by showcasing two attractive classes of configurations--configurations with judiciously chosen rotational transforms to avoid undesirable effects of low order resonances on the flux surface integrity and configurations with very small aspect ratios (∼2.5) that have excellent quasi-axisymmetry and low field ripples
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
International Nuclear Information System (INIS)
Kelly, J.L.; Rose, R.P.
1981-01-01
At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)
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 ...
Radiation (UV-IR) measurements in the TRIAM-1 tokamak
Energy Technology Data Exchange (ETDEWEB)
Nakamura, K; Nakamura, Y; Hiraki, N; Itoh, S [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics
1981-07-01
Scaling law is presented that the ratio P sub(R)/P sub(I) (P sub(R): radiation loss, P sub(I): input power) is proportional to the parameter a sup(3/2)/R (a: minor radius, R: major radius). The discussion is presented about the design of the apparatus for measuring the radiation (ultraviolet to infrared) from the TRIAM-1 tokamak plasma.
Tokamak startup using point-source dc helicity injection.
Battaglia, D J; Bongard, M W; Fonck, R J; Redd, A J; Sontag, A C
2009-06-05
Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.
Asymptotic stability boundaries of ballooning modes in circular tokamaks
International Nuclear Information System (INIS)
Chen, L.; Bondeson, A.; Chance, M.S.
1987-06-01
The model ballooning mode equation of Connor, Hastie, and Taylor for large-aspect-ratio circular tokamaks is analyzed in the limit of large pressure gradient, and corresponding expressions for stability boundaries are derived. In particular, it is found that for a fixed radial wave number, there exists an infinite sequence of unstable bands, and that minimizing over the radial wave numbers leads to asymptotic merging between the neighboring bands
Note for the Mirnov signal analysis in tokamaks
International Nuclear Information System (INIS)
Kikuchi, M.
1985-05-01
The relation between Mirnov coil signals and the current perturbation on the rational surface is examined analytically by using the approximate Green's function for the case of large aspect ratio circular tokamaks. Satellite island formation, phase modulation effect due to the poloidal variation of the field line pitch, and the shift effect of the plasma column with respect to the center of the vacuum chamber are examined. The detectability of these effects from Mirnov coil signals is discussed for TFTR
Closed expressions for the magnetic field of toroidal multipole configurations
International Nuclear Information System (INIS)
Sheffield, G.V.
1983-04-01
Closed analytic expressions for the vector potential and the magnetic field for the lower order toroidal multipoles are presented. These expressions can be applied in the study of tokamak plasma cross section shaping. An example of such an application is included. These expressions also allow the vacuum fields required for plasma equilibrium to be specified in a general form independent of a particular coil configuration
Software configuration management
International Nuclear Information System (INIS)
Arribas Peces, E.; Martin Faraldo, P.
1993-01-01
Software Configuration Management is directed towards identifying system configuration at specific points of its life cycle, so as to control changes to the configuration and to maintain the integrity and traceability of the configuration throughout its life. SCM functions and tasks are presented in the paper
Configuring the autism epidemic
DEFF Research Database (Denmark)
Seeberg, Jens; Christensen, Fie Lund Lindegaard
2017-01-01
Autism has been described as an epidemic, but this claim is contested and may point to an awareness epidemic, i.e. changes in the definition of what autism is and more attention being invested in diagnosis leading to a rise in registered cases. The sex ratio of children diagnosed with autism...... is skewed in favour of boys, and girls with autism tend to be diagnosed much later than boys. Building and further developing the notion of ‘configuration’ of epidemics, this article explores the configuration of autism in Denmark, with a particular focus on the health system and social support to families...... with children diagnosed with autism, seen from a parental perspective. The article points to diagnostic dynamics that contribute to explaining why girls with autism are not diagnosed as easily as boys. We unfold these dynamics through the analysis of a case of a Danish family with autism....
International Nuclear Information System (INIS)
Yamazaki, K.; Kawahata, K.; Ando, R.
1986-09-01
Disruptive behavior of magnetic field configuration leading to tokamak major disruption is found to be controlled by a mild ''mini-disruption'' which is induced by the compact external modular multipole-field coils with m = 3/n = 2 dominant helical field component in the JIPP T-IIU tokamak. This mini-disruption ergodizes the m = 2/n = 1 magnetic island quickly but mildly and then prevents the profile of electron temperature from flattening. This quick profile-reorganization is effective to avoid the two-step disruption (pre- and major disuptions) responsible for the chatastrophic current termination. (author)
Conceptualizing Embedded Configuration
DEFF Research Database (Denmark)
Oddsson, Gudmundur Valur; Hvam, Lars; Lysgaard, Ole
2006-01-01
and services. The general idea can be named embedded configuration. In this article we intend to conceptualize embedded configuration, what it is and is not. The difference between embedded configuration, sales configuration and embedded software is explained. We will look at what is needed to make embedded...... configuration systems. That will include requirements to product modelling techniques. An example with consumer electronics will illuminate the elements of embedded configuration in settings that most can relate to. The question of where embedded configuration would be relevant is discussed, and the current...
Development and Operational Experiences of the JT-60U Tokamak and Power Supplies
International Nuclear Information System (INIS)
Hosogane, N.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Neyatani, Y.; Horiike, H.; Sakurai, S.; Masaki, K.; Yamamoto, M.; Kodama, K.; Sasajima, T.; Terakado, T.; Ohmori, S.; Ohmori, Y.; Okano, J.
2002-01-01
The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-60 in order to obtain high plasma current, large plasma volume, and highly elongated divertor configurations. All components inside the toroidal magnetic field coils, such as vacuum vessel, poloidal magnetic field coils, divertor, etc., were modified. Various technologies and ideas were introduced to develop these components; for example, a multi-arc double skin wall structure for the vacuum vessel and a functional poloidal magnetic field coil system with taps for obtaining various plasma configurations. Furthermore, boron-carbide coated carbon fiber composite (CFC) tiles were used as divertor tiles to reduce erosion of carbon-base tiles. Later, a semiclosed divertor with pumps, for which cryo-panels originally used for NBI units were converted, was installed in the replacement of the open divertor. These development and operational results provide data for future tokamaks. Major failures experienced in the long operational period of JT-60U, such as water leakage from the toroidal magnetic field coil, fracture of carbon tiles, and breakdown of a filter capacitor, are described. As a maintenance issue for tokamaks using deuterium fueling gas, a method for reducing radiation exposure of in-vessel workers is described
Effect of recycling in the HL-1M tokamak
International Nuclear Information System (INIS)
Zheng Yongzhen
2004-01-01
Tokamak plasma discharge disruption at high density is investigated. The instability analysis on model indicates that the disruption is resulted from the energy loss arising from hydrogen recycling on the edge of the plasma. This energy loss could lead to a contraction of the current channel and the production of a disruptively unstable configuration. Using a simple model we shall investigate the implications of recycling for disruptions. The critical high-density n≤1.6 x 10 20 m -3 is reached in HL-1M. (author)
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
Multiple view fan beam polarimetry on Tokamak devices
International Nuclear Information System (INIS)
Geck, W.R.; Domier, C.W.; Luhmann, N.C.
1997-01-01
A polarimeter diagnostic is under development which utilizes several fan beams to accumulate line integrated Faraday rotation data in a Tokamak plasma. The utilization of a fan beam configuration over that of conventional vertical view polarimeter systems significantly reduces access requirements. The high angular separation inherent in a fan beam implementation increases plasma coverage and eliminates the necessity of assumed plasma symmetries to generate high quality current density profiles. Codes have been developed to generate these high-resolution two-dimensional images of the plasma current profile from data collected at arbitrary positions and viewing angles. copyright 1997 American Institute of Physics
Edge kink/ballooning mode stability in tokamaks with separatrix
International Nuclear Information System (INIS)
Medvedev, S Yu; Martynov, A A; Martin, Y R; Sauter, O; Villard, L
2006-01-01
Stability limits against external kink modes driven by large current density and pressure gradient values in the pedestal region are investigated for tokamak plasmas with separatrix. Stability diagrams for modes with different toroidal wave numbers under variations of pressure gradient and current density in the pedestal region are presented for several equilibrium configurations related to TCV. A scaling for the toroidal wave number of the most unstable mode is proposed. The influence of the plasma cross-section geometry on the stability limits is discussed
Coupling of tearing modes in tokamaks
International Nuclear Information System (INIS)
Finn, J.M.
1977-01-01
The simultaneous presence of tearing modes of different helical pitches leads to the destruction of magnetic surfaces, which has been suggested as the mechanism leading to the onset of the disruptive instability in tokamaks. For current profiles in which the m = 2 mode is unstable, but the m = 3 is stable, the coupling of the m = 3 to the m = 2 through the poloidal variation of the toroidal field can drive the m = 3 amplitude psi 3 to order psi 2 times the inverse aspect ratio. Detailed calculations, both analytical and numerical, have been performed for two models for the equilibrium and m = 2 mode structure. A slab model and incompressible m = 3 perturbations are assumed. The m = 3 amplitude increases with shear, up to a point, showing that as the current channel shrinks, overlap of resonances becomes more likely. The results also apply qualitatively to other m, m +- 1 interactions
Magnetic mapping of the TBR-1 tokamak
International Nuclear Information System (INIS)
Matta, Jose Antonio Sevidanes da.
1994-01-01
Axisymmetric hydromagnetic equilibria for an ideal conducting current fluid are described by means of an asymptotic expansion in powers of the inverse aspect ratio (ε α/R 0 ) that satisfies the Grad-Shafranov equation (equilibrium condition). The main profiles for these equilibria were computer. The effect of non-axisymmetric perturbations on the magnetic surfaces if the tokamak TBR-1 is investigated. The used method is able to show how magnetic field from external helical currents split the rational magnetic surface giving rise to the chain of magnetic islands. The Poincare map of the field line perturbed by resonant helical windings has been obtained numerically for the typical TBR-1 parameters. For increasing parameters of perturbation secondary resonances were observed, that transform the bound-state-like contours of a given island into similar structures of secondary magnetic islands. These results were used to determine the spectrum of the perturbation created by resonant helical windings. (author). 87 refs., 30 figs
Tearing mode analysis in tokamaks, revisited
International Nuclear Information System (INIS)
Nishimura, Y.; Callen, J.D.; Hegna, C.C.
1997-12-01
A new Δ' shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio (ε ≤ 0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of finite pressure term. Numerical results compare favorably with Furth et al. results. The effects of finite pressure, which are shown to decrease Δ', are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric element stabilizes the tearing mode significantly, even in a low β regime before the toroidal magnetic curvature effects come into play. Double tearing modes in toroidal geometries are examined as well. Furthermore, m ≥ 2 tearing mode stability criteria are compared with three dimensional initial value MHD simulation by the FAR code
Compact tokamak reactors part 2 (numerical results)
International Nuclear Information System (INIS)
Wiley, J.C.; Wootton, A.J.; Ross, D.W.
1996-01-01
The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced
Engineering parameters for four ignition TNS tokamak reactor systems
International Nuclear Information System (INIS)
Varljen, T.C.; Gibson, G.; French, J.W.; Heck, F.M.
1977-01-01
The ORNL/Westinghouse program for The Next Step (TNS) tokamak beyond TFTR has examined a large number of potential configurations for D-T burning ignition tokamak systems. An objective of this work has been to quantify the trade-offs associated with the assumption of certain plasma physics criteria and toroidal field coil technologies. Four tokamak system point designs are described, each representative of the TF coil technologies considered, to illustrate the engineering features associated with each concept. Point designs, such as the ones discussed herein, have been used to develop component size, performance and cost scaling relationships which have been incorporated in a digital computer code to facilitate an examination of the total design and cost impact of candidate design approaches. The point designs which are described are typical, however, they have not been individually optimized. The options are distinguished by the TF coil technology chosen and include: (1) a high field water-cooled copper TF system, (2) a moderate field NbTi superconducting TF system, (3) a high field Nb 3 Sn superconducting TF system, and (4) a high field hybrid TF system with outer NbTi superconducting windings and inner water-cooled copper windings. Descriptions are provided for the major device components and all major support systems including power supplies, vacuum systems, fuel systems, heat transport and facility systems
Energy losses on tokamak startup
International Nuclear Information System (INIS)
Murray, J.G.; Rothe, K.E.; Bronner, G.
1983-01-01
During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells
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
Tokamak impurity-control techniques
International Nuclear Information System (INIS)
Schmidt, J.A.
1980-01-01
A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)
An enhanced tokamak startup model
Goswami, Rajiv; Artaud, Jean-François
2017-01-01
The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.
Multimegawatt neutral beams for tokamaks
International Nuclear Information System (INIS)
Kunkel, W.B.
1979-03-01
Most of the large magnetic confinement experiments today and in the near future use high-power neutral-beam injectors to heat the plasma. This review briefly describes this remarkable technique and summarizes recent results as well as near term expectations. Progress has been so encouraging that it seems probable that tokamaks will achieve scientific breakeven before 1990
Joint research using small tokamaks
Czech Academy of Sciences Publication Activity Database
Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.
2005-01-01
Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005
Simulations of edge and scrape off layer turbulence in mega ampere spherical tokamak plasmas
DEFF Research Database (Denmark)
Militello, F; Fundamenski, W; Naulin, Volker
2012-01-01
The L-mode interchange turbulence in the edge and scrape-off-layer (SOL) of the tight aspect ratio tokamak MAST is investigated numerically. The dynamics of the boundary plasma are studied using the 2D drift-fluid code ESEL, which has previously shown good agreement with large aspect ratio machin...
On steady poloidal and toroidal flows in tokamak plasmas
International Nuclear Information System (INIS)
McClements, K. G.; Hole, M. J.
2010-01-01
The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B θ /B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B θ /B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.
The conceptual design for the modification of HL-2A tokamak
International Nuclear Information System (INIS)
Dequan Liu
2006-01-01
The medium-sized tokamak HL-2A, based on the former ASDEX's main components has approached its rating operational parameters with I p = 450 kA and B T =2.8 T so far. The HL-2A was originally designed to operate under an axisymmetrical double-null configuration by two triplets of shaping coils located in the vacuum vessel. Because the shaping coils were set very close to the plasma column, the nulls are normally fixed and the plasma has a nearly circular cross-section. It is difficult to increase the parameters further and obtain a preferable plasma shape with certain values of triangularity and elongation. Due to the defects on upper shaping coils' electrical insulation, the HL-2A is presently just operated with lower single null divertor configuration. Additionally, the experiment has to occasionally terminate due to incidental damages of the weak parts. With the experimental progress on HL-2A, the plasma stored energy and auxiliary heating power is being increased, a more efficient, compact and tight divertor is needed. Based on present status of HL-2A tokamak and for satisfying the requirements of the experiment, the modification of HL-2A tokamak device will be carried out. The modification design of HL-2A intents to obtain optimized plasma parameters, e.g., aspect ratio, I p , b, plasma volume, certain plasma shaping with preferable elongation and triangularity, a stable vacuum vessel easy to be maintained and an advanced divertor. The primary consideration of the modification is to take all of the shaping coils out of the vessel to enlarge the plasma volume. The elongated and triangular cross section with double null of plasma is to be obtained by composite distribution of eight sets of poloidal coils. Because all poloidal coils will be located between the TF coil and the vessel, the vacuum vessel will be remade with a smaller size to make rooms for the poloidal coil system. With an optimized operation mode, the volt-second of the plasma may be increased a
Current drive by Alfven waves in elongated cross section tokamak
International Nuclear Information System (INIS)
Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A.; Assis, A.S. de
1997-01-01
Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves 9GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory
Current drive by Alfven waves in elongated cross section tokamak
Energy Technology Data Exchange (ETDEWEB)
Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Fisica; Assis, A.S. de [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Inst. de Fisica
1997-12-31
Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves (GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory
Including collisions in gyrokinetic tokamak and stellarator simulations
International Nuclear Information System (INIS)
Kauffmann, Karla
2012-01-01
Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the δf method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal ITGs and
Including collisions in gyrokinetic tokamak and stellarator simulations
Energy Technology Data Exchange (ETDEWEB)
Kauffmann, Karla
2012-04-10
Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal
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.
Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation
International Nuclear Information System (INIS)
Riedel, K.S.
1989-01-01
This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies
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)
Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.
1975-01-01
A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR
Shield design for next-generation, low-neutron-fluence, superconducting tokamaks
International Nuclear Information System (INIS)
Lee, V.D.; Gohar, Y.
1985-01-01
A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of builtup, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components
Shield design for next-generation, low-neutron-fluence, superconducting tokamaks
International Nuclear Information System (INIS)
Lee, V.D.; Gohar, Y.
1985-01-01
A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components
A comprehensive theory of the equilibria in a tokamak and a reversed field pinch
International Nuclear Information System (INIS)
Chiyoda, Katsuji
1996-01-01
The equilibrium configuration of a tokamak is analysed by the equilibrium equations derived for analysing a reversed field pinch (RFP). The expressions of the magnetic field and the toroidal shift in the internal plasma region and the external vacuum region are obtained. The expressions in the vacuum region become the Shafranov's expressions, when the plasma-center coordinates is used. Discontinuities of the equilibrium quantities are considered. It is concluded that the equilibrium equations are applicable also to the tokamak plasma and that the difference of the equilibria between the tokamak and the RFP stems from the choices of the pressure and the toroidal current function. A feature of our theory is that any ordering to the safety factor is not imposed. (author)
International Nuclear Information System (INIS)
Burgess, M.
1994-06-01
A high level description language is presented for the purpose of automatically configuring large heterogeneous networked unix environments, based on class-oriented abstractions. The configuration engine is portable and easily extensible
Neoclassical dissipation and resistive wall modes in tokamaks
International Nuclear Information System (INIS)
Shaing, K.C.
2004-01-01
It is shown that the critical toroidal plasma flow speed that is required to stabilize the resistive wall mode in tokamaks is reduced by a factor of the order of B/B θ or of 1.265ε 3sol4 B/B θ depending on the plasma parameters when the perturbed neoclassical viscosity driven current is taken into account. Here, B is the magnetic field strength, B θ is the poloidal magnetic field strength, and ε is the inverse aspect ratio. This effect is illustrated using an existing model for the resistive wall modes by including the neoclassical dissipation in the derivation of the dispersion relation. The derivation is based on fluid equations with the plasma viscosity, calculated using kinetic equation, as the closure. The reduction of the critical toroidal speed is a consequence of the parallel (to the magnetic field B) momentum equation when neoclassical viscosity becomes important. The results are compared with experimental observations in tokamaks
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/{{\
Multi-channel bolometer system on JFT-2M tokamak
International Nuclear Information System (INIS)
Tamai, Hiroshi; Maeno, Masaki; Matsuda, Toshiaki; Matoba, Tohru
1988-07-01
Multi-channel bolometer system is designed and installed to observe the radiation profile on JFT-2M tokamak. Sensor head is made of Thinistor, which is a kind of semiconductor, because it has the advantage of higher sensitivity of about one order of magnitude than the conventional metal foil bolometer and is suitable for the profile measurement in which the signal from the plasma is relatively small. The response and cooling characteristics of the bolometer sensor are suitable for the condition of JFT-2M tokamak plasma. Low noise circuit of bridge and differentiator is developed to optimize the signal to noise ratio in the JFT-2M operating condition. With use of the bolometer system, the radiation profile in joule heating plasma as well as additional heating plasma especially in H-mode plasma is successfully observed. (author)
Conceptual design of a Tokamak hybrid power reactor (THPR)
International Nuclear Information System (INIS)
Matsuoka, F.; Imamura, Y.; Inoue, M.; Asami, N.; Kasai, M.; Yanagisawa, I.; Ida, T.; Takuma, T.; Yamaji, K.; Akita, S.
1987-01-01
A conceptual design of a fusion-fission hybrid tokamak reactor has been carried out to investigate the engineering feasibility and promising scale of a commercial hybrid reactor power plant. A tokamak fusion driver based on the recent plasma scaling law is introduced in this design study. The major parameters and features of the reactor are R=6.06 m, a=1.66 m, Ip=11.8 MA, Pf=668 MW, double null divertor plasma and steady state burning with RF current drive. The fusion power has been determined with medium energy multiplication in the blanket so as to relieve thermal design problems and produce electric power around 1000 MW. Uranium silicide is used for the fast fission blanket material to promise good nuclear performance. The coolant of the blanket is FLIBE and the tritium breeding blanket material is Li 2 O ceramics providing breeding ratio above unity
MICROCONTROLLER PIN CONFIGURATION TOOL
Bhaskar Joshi; F. Mohammed Rizwan; Dr. Rajashree Shettar
2012-01-01
Configuring the micro controller with large number of pins is tedious. Latest Infine on microcontroller contains more than 200 pins and each pin has classes of signals. Therefore the complexity of the microcontroller is growing. It evolves looking into thousands of pages of user manual. For a user it will take days to configure the microcontroller with the peripherals. We need an automated tool to configure the microcontroller so that the user can configure the microcontroller without having ...
Operational Dynamic Configuration Analysis
Lai, Chok Fung; Zelinski, Shannon
2010-01-01
Sectors may combine or split within areas of specialization in response to changing traffic patterns. This method of managing capacity and controller workload could be made more flexible by dynamically modifying sector boundaries. Much work has been done on methods for dynamically creating new sector boundaries [1-5]. Many assessments of dynamic configuration methods assume the current day baseline configuration remains fixed [6-7]. A challenging question is how to select a dynamic configuration baseline to assess potential benefits of proposed dynamic configuration concepts. Bloem used operational sector reconfigurations as a baseline [8]. The main difficulty is that operational reconfiguration data is noisy. Reconfigurations often occur frequently to accommodate staff training or breaks, or to complete a more complicated reconfiguration through a rapid sequence of simpler reconfigurations. Gupta quantified a few aspects of airspace boundary changes from this data [9]. Most of these metrics are unique to sector combining operations and not applicable to more flexible dynamic configuration concepts. To better understand what sort of reconfigurations are acceptable or beneficial, more configuration change metrics should be developed and their distribution in current practice should be computed. This paper proposes a method to select a simple sequence of configurations among operational configurations to serve as a dynamic configuration baseline for future dynamic configuration concept assessments. New configuration change metrics are applied to the operational data to establish current day thresholds for these metrics. These thresholds are then corroborated, refined, or dismissed based on airspace practitioner feedback. The dynamic configuration baseline selection method uses a k-means clustering algorithm to select the sequence of configurations and trigger times from a given day of operational sector combination data. The clustering algorithm selects a simplified
Microwave Tokamak Experiment: Overview and status
International Nuclear Information System (INIS)
1990-05-01
The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs
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
Tokamak transmutation of (nuclear) waste (TTW): Parametric studies
International Nuclear Information System (INIS)
Cheng, E.T.; Krakowski, R.A.; Peng, Y.K.M.
1994-01-01
Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low-aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses
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.
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)
Commissioning and preliminary operation of HL-2A tokamak
International Nuclear Information System (INIS)
Liu Dequan; Liu Yong; Yan Jianchen; Cao Zeng; Yang Qingwei; Zhou Caipin; Li Xiaodong
2005-01-01
HL-2A is a divertor tokamak located at new site of Southwestern Institute of Physics (SWIP), Chengdu, China. It can be operated in double-null and single-null divertor with closed configurations. The effects of the divertor on impurity behaviors, MHD instabilities, transport, wall conditioning and divertor physics are key issues to be studied during the first step operation on HL-2A. Preliminary experiments with limiter and single null divertor configurations were carried out in 2003. Main parameters achieved are the plasma current 168 KA, duration time of 920 ms and plasma linear-averaged density of 1.7 x 10 19 m -3 . The impurities, especially of low Z, are clearly decreased during the divertor experiments
International Nuclear Information System (INIS)
Nakamura, Yukiharu; Nishio, Satoshi; Yoshino, Ryuji; Kessel, C.E.; Jardin, S.C.
1996-01-01
The dynamic behavior of vertical displacement events (VDEs) during a disruption and acceleration mechanisms that govern VDEs in the ITER-EDA tokamak are investigated using the Tokamak Simulation Code. A sudden plasma pressure drop (β p collapse) does not accelerate VDEs for the ITER tokamak. The geometry of the ITER resistive shell is shown to be suitable for preventing a β p collapse-induced VDE, because the magnetic field decay n-index after the β p collapse does not considerably degrade. On the other hand, it is shown that the plasma current quench (I p quench) following the energy quench can accelerate VDEs due to the vertical imbalance of the attractive force arising from the up-down asymmetric shell. The vertical location of the neutral point where the I p quench-induced VDE almost disappears is found to lie at ∼22 cm below the plasma magnetic axis of the nominal equilibrium (Z = 1.44 m). An upward and moderate I p quench-induced VDE can be expected for the nominal configuration in the ITER-EDA tokamak. It is shown that the ITER tokamak has an advantage of avoiding the fatal damage of the complicated structures of the bottom-divertor. (author)
The timing system on the J-TEXT tokamak
International Nuclear Information System (INIS)
Zheng, Wei; Zhang, Ming; Zhuang, Ge; Ding, Tonghai; Huang, Fuqiang; Shan, Lingjie
2014-01-01
Highlights: •The timing system achieved tree structured timing network with only one type of timing module. •This system is integrated into J-TEXT COADC which is an EPICS based control system. •This system handles multiple timing sequences and events. •This system has been deployed on J-TEXT and working properly in daily experiments. -- Abstract: This paper describes the timing system designed to control the operation time-sequence and to generate clocks for various sub-systems on J-TEXT tokamak. The J-TEXT timing system is organized as a distributed system which is connected by a tree-structured optical fiber network. It can generate delayed triggers and gate signals (0 μs–4000 s), while providing reference clocks for other sub-systems. Besides, it provides event handling and timestamping functions. It is integrated into the J-TEXT Control, Data Access and Communication (J-TEXT CODAC) system, and it can be monitored and configured by Experimental Physics and Industrial Control System (EPICS). The configuration of this system including tree-structured network is managed in XML files by dedicated management software. This system has already been deployed on J-TEXT tokamak and it is serving J-TEXT in daily experiments
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)
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.
Flux driven turbulence in tokamaks
International Nuclear Information System (INIS)
Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.
1999-01-01
This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)
Options for an ignited tokamak
International Nuclear Information System (INIS)
Sheffield, J.
1984-02-01
It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon β/sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed
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
Starfire: a commercial tokamak reactor
International Nuclear Information System (INIS)
Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.
1979-01-01
The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE
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
Demonstration tokamak-power-plant study (DEMO)
International Nuclear Information System (INIS)
1982-09-01
A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material
Magnet design approach for pulsed tokamak reactors
International Nuclear Information System (INIS)
Kim, S.H.; Evans, K. Jr.; Ehst, D.A.
1983-12-01
A choice of various operating modes of a tokamak reactor will have considerable impact on the fatigue lives and cost of ohmic heating (OH), equilibrium field (EF), and toroidal field (TF) coils. OH AND EF coil requirements and their costs, as well as the effects of the fringing fields of the EF coils on the TF coils, have been studied under cyclic operation in the range of N = 10 2 to 10 6 cycles, spanning the range from a noninductively driven reactor (STARFIRE) to a conventional ohmically driven reactor. For a reference design of TF coils the design of the central OH solenoid has been studied as a function of its maximum field, B/sup OH/. Increasing requirements for structural support lead to only negligible increases in volt-seconds for B/sup OH/ greater than or equal to 10.0 T. Fatigue failure of the OH coil is not a concern for N less than or equal to 10 5 ; for N approx. 10 6 fatigue limits the strain to small values, resulting in small increases in structural requirements and modest decreases in volt-seconds. Should noninductive current drive be achievable we note that this not only eliminates the OH coil, but it also permits EF coil placement in the inboard region, which facilitates the creation of highly shaped plasma cross sections (large triangularity, or bean-shaped equilibria). We have computed the stored energy, coil configuration and fringing fields for a number of EF coil design options
Magnetic island formation in tokamaks
International Nuclear Information System (INIS)
Yoshikawa, S.
1989-04-01
The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs
Equilibrium Reconstruction in EAST Tokamak
International Nuclear Information System (INIS)
Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.
2009-01-01
Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)
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.)
Runaway electrons during tokamak startup
International Nuclear Information System (INIS)
Sharma, A.S.; Jayakumar, R.
1988-01-01
Runaway electrons significantly affect the plasma and impurity evolution during tokamak startup. During its rise, a runaway pulse stores magnetic flux inductively; this is then released during the decay phase of the runaway pulse. This process affects plasma formation, current initiation and current buildup. Because of their relativistic velocities the runaway electrons have higher ionization and excitation rates than the plasma electrons. This leads to a significant modification of the impurity behaviour and consequently the plasma evolution. (author). 20 refs, 8 figs
Minimum scaling laws in tokamaks
International Nuclear Information System (INIS)
Zhang, Y.Z.; Mahajan, S.M.
1986-10-01
Scaling laws governing anomalous electron transport in tokamaks with ohmic and/or auxiliary heating are derived using renormalized Vlasov-Ampere equations for low frequency electromagnetic microturbulence. It is also shown that for pure auxiliary heating (or when auxiliary heating power far exceeds the ohmic power), the energy confinement time scales as tau/sub E/ ∼ P/sub inj//sup -1/3/, where P/sub inj/ is the injected power
Gyrosheath near the tokamak edge
International Nuclear Information System (INIS)
Hazeltine, R.D.; Xiao, H.; Valanju, P.M.
1993-03-01
A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results
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
Shear Alfven waves in tokamaks
International Nuclear Information System (INIS)
Kieras, C.E.
1982-12-01
Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma
Discharge cleaning for a tokamak
International Nuclear Information System (INIS)
Ishii, Shigeyuki
1983-01-01
Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)
Magnetic confinement experiment. I: Tokamaks
International Nuclear Information System (INIS)
Goldston, R.J.
1995-08-01
Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices
Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade
International Nuclear Information System (INIS)
Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.
2015-01-01
Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)
Control of tokamak plasma current and equilibrium with hybrid poloidal field coils
International Nuclear Information System (INIS)
Shimada, Ryuichi
1982-01-01
A control method with hybrid poloidal field system is considered, which comprehensively implements the control of plasma equilibrium and plasma current, those have been treated independently in Tokamak divices. Tokamak equilibrium requires the condition that the magnetic flux function value on plasma surface must be constant. From this, the current to be supplied to each coil is determined. Therefore, each coil current is the resultant of the component related to plasma current excitation and the component required for holding equilibrium. Here, it is intended to show a method by which the current to be supplied to each coil can easily be calculated by the introduction of hybrid control matrix. The text first considers the equilibrium of axi-symmetrical plasma and the equilibrium magnetic field outside plasma, next describes the determination of current using the above hybrid control matrix, and indicates an example of controlling Tokamak plasma current and equilibrium by the hybrid poloidal field coils. It also shows that the excitation of plasma current and the maintenance of plasma equilibrium can basically be available with a single power supply by the appropriate selection of the number of turns of each coil. These considerations determine the basic system configuration as well as decrease the installed capacity of power source for the poloidal field of a Tokamak fusion reactor. Finally, the actual configuration of the power source for hybrid poloidal field coils is shown for the above system. (Wakatsuki, Y.)
HLT configuration management system
Daponte, Vincenzo
2015-01-01
The CMS High Level Trigger (HLT) is implemented running a streamlined version of the CMS offline reconstruction software running on thousands of CPUs. The CMS software is written mostly in C++, using Python as its configuration language through an embedded CPython interpreter. The configuration of each process is made up of hundreds of modules, organized in sequences and paths. As an example, the HLT configurations used for 2011 data taking comprised over 2200 different modules, organized in more than 400 independent trigger paths. The complexity of the HLT configurations and the large number of configuration produced require the design of a suitable data management system. The present work describes the designed solution to manage the considerable number of configurations developed and to assist the editing of new configurations. The system is required to be remotely accessible and OS-independent as well as easly maintainable easy to use. To meet these requirements a three-layers architecture has been choose...
Configuration development and structural assessment of the FEDC ignitor concept
International Nuclear Information System (INIS)
Brown, T.G.; Lee, V.D.; Mayhall, J.A.
1986-01-01
The Fusion Engineering Design Center (FEDC) performed a design study for a compact ignition tokamak. The ignitor concept has two unique features. First, the throat of the copper plate toroidal field (TF) coils is preloaded in the vertical direction to minimize the stress levels in the copper. Second, the net inward radial TF coil forces are balanced by a combination of wedging on the adjacent faces of the TF coils and by bucking against the ohmic heating solenoid coils in the bore of the tokamak. Later ignitor concepts eliminated the wedging reactions. Both of these features inherently reduce the radial build of the tokamak device. The FEDC version of ignitor incorporates both of these unique features packaged in a different configurational arrangement. The FEDC ignitor features a totally external preload system. The preload is applied directly to the inner leg of the TF coils in the vertical direction only. Horizontal rings are utilized only to react inplane TF coil forces and are not part of the preload system. Modular quadrants of TF coil encasements are utilized to vastly simplify device assembly methods. This paper presents the configuration and a structural assessment of the FEDC ignitor concept
International Nuclear Information System (INIS)
Brunner, S.
1997-08-01
Ion temperature gradient (ITG)-related instabilities are studied in tokamak-like plasmas with the help of a new global eigenvalue code. Ions are modelled in the frame of gyrokinetic theory so that finite Larmor radius effects of these particles are retained to all orders. Non-adiabatic trapped electron dynamics is taken into account through the bounce-averaged drift kinetic equation. Assuming electrostatic perturbations, the system is closed with the quasineutrality relation. Practical methods are presented which make this global approach feasible. These include a non-standard wave decomposition compatible with the curved geometry as well as adapting an efficient root finding algorithm for computing the unstable spectrum. These techniques are applied to a low pressure configuration given by a large aspect ratio torus with circular, concentric magnetic surfaces. Simulations from a linear, time evolution, particle in cell code provide a useful benchmark. Comparisons with local ballooning calculations for different parameter scans enable further validation while illustrating the limits of that representation at low toroidal wave numbers or for non-interchange-like instabilities. The stabilizing effect of negative magnetic shear is also considered, in which case the global results show not only an attenuation of the growth rate but also a reduction of the radial extent induced by a transition from the toroidal- to the slab-ITG mode. Contributions of trapped electrons to the ITG instability as well as the possible coupling to the trapped electron mode are clearly brought to the fore. (author) figs., tabs., 69 refs
Test results on systems developed for SST-1 tokamak
International Nuclear Information System (INIS)
Bora, D.
2003-01-01
Steady state Superconducting Tokamak (SST-1) is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation (κ) and triangularity Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. A NbTi based cable-in-conduit conductor (CICC) has been fabricated by M/S Hitachi Cables Ltd., Japan under specification and supervision of IPR. The suitability of this CICC for the SST-1 magnets has been validated through test carried out on a model coil (MC) wound from this CICC. Toroidal and poloidal SC magnets have been fabricated and factory acceptance tests have been performed. SC magnets require liquid helium (LHe) cooled current leads, electrical isolators at LHe temperature, superconducting bus bars and LHe transfer lines. Full scale prototypes of these have been developed and tested successfully. SC magnets will be cooled to 4.5K by forced flow of supercritical Helium through the CICC. A 1 kW grade liquefier/refrigerator has been installed and is in final stages of commissioning at IPR. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors having ports and 16 rings with D-shaped cross-section. To establish the fabrication methodology for this, a full scale proto-type of the vessel with two vessel sectors and three rings has been fabricated and tested successfully. Based on this the fabrication of the vessel sectors and rings is in final stage of fabrication. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. SST-1 will have three different high power radio frequency (RF) systems to additionally heat and non-inductively drive plasma current to sustain the plasma in steady state for a duration of up to 1000 sec. Ion cyclotron resonance frequency (ICRF) and electron cyclotron resonance frequency (ECRF) systems will primarily be
System engineering and configuration management in ITER
International Nuclear Information System (INIS)
Chiocchio, S.; Martin, E.; Barabaschi, P.; Bartels, Hans Werner; How, J.; Spears, W.
2007-01-01
The construction of ITER will represent a major challenge for the fusion community at large, because of the intrinsic complexity of the tokamak design, the large number of different systems which are all essential for its operation, the worldwide distribution of the design activities and the unusual procurement scheme based on a combination of in-kind and directly funded deliverables. A key requirement for the success of such a large project is that a systematic approach to ensure the consistency of the design with the required performance is adopted. Also, effective project management methods, tools and working practices must be deployed to facilitate the communication and collaboration among the institutions and industries involved in the project. The authors have been involved in the definition and practical implementation of the design integration and configuration control structure inside ITER and in the system engineering process during the selection and optimization of the machine configuration. In parallel, they have assessed design, drawing and documentation management software to be used for the construction phase. Here, they describe the experience gained in recent years, explain the drivers behind the selection of the documents and drawings management systems, and illustrate the scope and issues of the configuration management activities to ensure the congruence of the design, to control and track the design changes and to manage the interfaces among the ITER systems
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)
Zonal flows in tokamaks with anisotropic pressure
International Nuclear Information System (INIS)
Ren, Haijun
2014-01-01
Zonal flows (ZFs) in a tokamak plasma with anisotropic pressure are investigated. The dynamics of perpendicular and parallel pressures are determined by the Chew-Goldberger-Low double equations and low-β condition is adopted, where β is the ratio of plasma pressure to the magnetic field pressure. The dispersion relation is analytically derived and illustrates two branches of ZFs. The low frequency zonal flow (LFZF) branch becomes unstable when χ, the ratio of the perpendicular pressure to the parallel one, is greater than a threshold value χ c , which is about 3.8. In the stable region, its frequency increases first and then decreases with increasing χ. For χ = 1, the frequency of LFZF agrees well with the experimental observation. For the instability, the growth rate of LFZF increases with χ. The geodesic acoustic mode branch is shown to be always stable with a frequency increasing with χ. The safety factor is shown to diminish the frequencies of both branches or the growth rate of LFZF
Economic evaluation of tokamak power plants
International Nuclear Information System (INIS)
Reid, R.L.; Steiner, D.
1977-01-01
This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented
Simulation of a major tokamak disruption
International Nuclear Information System (INIS)
White, R.B.; Monticello, D.A.; Rosenbluth, M.N.
1977-08-01
It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption
Diagnostics for the Rijnhuizen Tokamak Project
Donne, A. J. H.
1994-01-01
The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel
Mercier criterion for high-β tokamaks
International Nuclear Information System (INIS)
Galvao, R.M.O.
1984-01-01
An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt
The ARIES-I tokamak reactor study
International Nuclear Information System (INIS)
1991-01-01
This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations
Software configuration management
Keyes, Jessica
2004-01-01
Software Configuration Management discusses the framework from a standards viewpoint, using the original DoD MIL-STD-973 and EIA-649 standards to describe the elements of configuration management within a software engineering perspective. Divided into two parts, the first section is composed of 14 chapters that explain every facet of configuration management related to software engineering. The second section consists of 25 appendices that contain many valuable real world CM templates.
Summary report on tokamak confinement experiments
International Nuclear Information System (INIS)
1982-03-01
There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users
International Nuclear Information System (INIS)
Alsaed, A.
2004-01-01
''The Disposal Criticality Analysis Methodology Topical Report'' prescribes an approach to the methodology for performing postclosure criticality analyses within the monitored geologic repository at Yucca Mountain, Nevada. An essential component of the methodology is the ''Configuration Generator Model for In-Package Criticality'' that provides a tool to evaluate the probabilities of degraded configurations achieving a critical state. The configuration generator model is a risk-informed, performance-based process for evaluating the criticality potential of degraded configurations in the monitored geologic repository. The method uses event tree methods to define configuration classes derived from criticality scenarios and to identify configuration class characteristics (parameters, ranges, etc.). The probabilities of achieving the various configuration classes are derived in part from probability density functions for degradation parameters. The NRC has issued ''Safety Evaluation Report for Disposal Criticality Analysis Methodology Topical Report, Revision 0''. That report contained 28 open items that required resolution through additional documentation. Of the 28 open items, numbers 5, 6, 9, 10, 18, and 19 were concerned with a previously proposed software approach to the configuration generator methodology and, in particular, the k eff regression analysis associated with the methodology. However, the use of a k eff regression analysis is not part of the current configuration generator methodology and, thus, the referenced open items are no longer considered applicable and will not be further addressed
Ansible configuration management
Hall, Daniel
2013-01-01
Ansible Configuration Management"" is a step-by-step tutorial that teaches the use of Ansible for configuring Linux machines.This book is intended for anyone looking to understand the basics of Ansible. It is expected that you will have some experience of how to set up and configure Linux machines. In parts of the book we cover configuration files of BIND, MySQL, and other Linux daemons, therefore a working knowledge of these would be helpful but are certainly not required.
Configuration management at NEK
International Nuclear Information System (INIS)
Podhraski, M.
1999-01-01
Configuration Management (CM) objectives at NEK are to ensure consistency between Design Requirements, Physical Plant Configuration and Configuration Information. Software applications, supporting Design Change, Work Control and Document Control Processes, are integrated in one module-oriented Management Information System (MIS). Master Equipment Component List (MECL) database is central MIS module. Through a combination of centralized database and process migrated activities it is ensured that the CM principles and requirements (accurate, current design data matching plant's physical configuration while complying to applicable requirements), are followed and fulfilled.(author)
Study on ECH-assisted start-up using trapped particle configuration in KSTAR and application to ITER
Lee, Jeongwon; Kim, Jayhyun; An, YoungHwa; Yoo, Min-Gu; Hwang, Y. S.; Na, Yong-Su
2017-12-01
ECH-assisted start-up using trapped particle configuration (TPC) is firstly studied in a superconducting, conventional tokamak, KSTAR. First, improved and efficient start-up using TPC than conventional field null configuration (FNC) is achieved by enhanced pre-ionization plasma quality. TPC shows the broader operation window in terms of the poloidal field quality and the deuterium prefill pressure than that of FNC. Surprisingly the particle trapping enhances the plasma start-up performance even with much lower particle trapping ratio than that of spherical torus. Reliability of TPC with low trapping ratio is investigated by 0D plasma evolution code, TECHP0D. Second, the characteristics of TPC start-up are explored with experiments and modellings. Two kind of start-up failure conditions are identified by the magnetic pitch and prefill pressure scan experiments: (i) low ionization rate; and (ii) low density condition. These experimental observation has consistency with TECHP0D modeling result. Finally, reliable ITER-relevant low toroidal electric field start-up using TPC is achieved with drastically improved success rate of start-up in KSTAR. In this experiments, validity of the ITER-like toroidally inclined ECH/ECCD injection is also expected. Time dependent ITER start-up scenario using TPC is proposed and the superiority of it than FNC is discussed.
Engineering and thermal-hydraulic design of water cooled PFC for SST-1 tokamak
International Nuclear Information System (INIS)
Paritosh Chaudhuri; Santra, P.; Rabi Prakash, N.; Khirwadkar, S.; Arun Prakash, A.; Ramash, G.; Dubey, S.; Chenna Reddy, D.; Saxena, Y.C.
2005-01-01
Full text of publication follows: Steady state Superconducting Tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. It is a large aspect ratio tokamak with a major radius of 1.1 m and minor radius of 0.20 m. SST-1 is designed for plasma discharge duration of ∼1000 seconds to obtain fully steady state plasma with total input power up to 1.0 MW. First Wall or Plasma Facing Components (PFC) is one or the major sub-systems of SST-1 tokamak consisting of divertors, passive stabilizers, baffles, and poloidal limiters are designed to be compatible for steady state operation. All the PFC has the same basic configuration: graphite tiles are mechanically attached to a back plate made of high strength copper alloy, and SS tubes are embedded in the groove made in the back plate. Same tube will be used for cooling during plasma operation and baking during wall conditioning. The main consideration in the design of the PFC is the steady state heat removal of up to 1 MW/m 2 . In addition to remove high heat fluxes, the PFC are also designed to be compatible for high temperature baking at 350 deg. C. Water was chosen as the coolant because of its appropriate thermal properties, and while baking, hot nitrogen gas would flow through these tubes to bake the PFC at high temperature. Extensive studies, involving different flow parameters and various cooling layouts, has been done to select the final cooling parameters and layout, compatible for cooling and baking. During steady state operation, divertor and passive stabilizer heat loads are expected to be 0.6 and 0.25 MW/m 2 . The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Since the tile must be mechanically attached to the back plate (heat sink), the fitting technique must provide the highest mechanical stress so that thermal transfer efficiency is maximized. Proper brazing of cooling tube on the copper back
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.
Ion orbit loss and pedestal width of H-mode tokamak plasmas in limiter geometry
International Nuclear Information System (INIS)
Xiao Xiaotao; Liu Lei; Zhang Xiaodong; Wang Shaojie
2011-01-01
A simple analytical model is proposed to analyze the effects of ion orbit loss on the edge radial electric field in a tokamak with limiter configuration. The analytically predicted edge radial electric field is consistent with the H-mode experiments, including the width, the magnitude, and the well-like shape. This model provides an explanation to the H-mode pedestal structure. Scaling of the pedestal width based on this model is proposed.
Energy Technology Data Exchange (ETDEWEB)
Valencia A, R
2006-07-01
spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow discharge plasma, created in a calibrated mixture of methane-hydrogen during the hydrogenated amorphous carbon film deposit on the vessel wall of Novillo tokamak, were determined by mass spectrometry. By way of measuring the emission lines of the carbon and oxygen impurities in intense discharges, the time required by the plasma to interact with the wall was estimated. In addition to it, the temporal conduct of the emission line intensity of these impurities was observed by means of an intensified CCD detector. Once an {approx} 10 % of helium was introduced in the operating gas of the tokamak discharges, a 25-42 eV time variation of the electron temperature was measured using the intensity ratio technique. (Author)
Electron and impurity transport studies in the TCV Tokamak
Energy Technology Data Exchange (ETDEWEB)
Wagner, D.
2013-05-15
In this thesis electron and impurity transport are studied in the Tokamak à Configuration Variable (TCV) located at CRPP-EPFL in Lausanne. Understanding particle transport is primordial for future nuclear fusion power plants. Modeling of experiments in many specific plasma scenarios can help to understand the common elements of the physics at play and to interpret apparently contradictory experiments on the same machine and across different machines. The first part of this thesis deals with electron transport in TCV high confinement mode plasmas. It was observed that the electron density profile in these plasmas flatten when intense electron heating is applied, in contrast to observations on other machines where the increase of the profile peakedness was reported. It is shown with quasi-linear gyrokinetic simulations that this effect, usually interpreted as collisionality dependence, stems from the combined effect of many plasma parameters. The influence of the collisionality, electron to ion temperature ratio, the ratio of temperature gradients, and the Ware-pinch are studied with detailed parameter scans. It is shown that the complex interdependence of the various plasma parameters is greatly simplified when the simulation results are interpreted as a function of the average frequency of the main modes contributing to radial transport. In this way the model is able to explain the experimental results. It was also shown that the same basic understanding is at play in L-modes, H-modes and electron internal transport barriers. The second part of the thesis is devoted to impurity transport. A multi-purpose gas injection system is developed, commissioned and calibrated. It is shown that the system is capable of massive gas injections to provoke disruptions and delivering small puffs of gaseous impurities for perturbative transport experiments. This flexible tool is exploited in a series of impurity transport measurements with argon and neon injections. The impurities
Electron and impurity transport studies in the TCV Tokamak
International Nuclear Information System (INIS)
Wagner, D.
2013-05-01
In this thesis electron and impurity transport are studied in the Tokamak à Configuration Variable (TCV) located at CRPP-EPFL in Lausanne. Understanding particle transport is primordial for future nuclear fusion power plants. Modeling of experiments in many specific plasma scenarios can help to understand the common elements of the physics at play and to interpret apparently contradictory experiments on the same machine and across different machines. The first part of this thesis deals with electron transport in TCV high confinement mode plasmas. It was observed that the electron density profile in these plasmas flatten when intense electron heating is applied, in contrast to observations on other machines where the increase of the profile peakedness was reported. It is shown with quasi-linear gyrokinetic simulations that this effect, usually interpreted as collisionality dependence, stems from the combined effect of many plasma parameters. The influence of the collisionality, electron to ion temperature ratio, the ratio of temperature gradients, and the Ware-pinch are studied with detailed parameter scans. It is shown that the complex interdependence of the various plasma parameters is greatly simplified when the simulation results are interpreted as a function of the average frequency of the main modes contributing to radial transport. In this way the model is able to explain the experimental results. It was also shown that the same basic understanding is at play in L-modes, H-modes and electron internal transport barriers. The second part of the thesis is devoted to impurity transport. A multi-purpose gas injection system is developed, commissioned and calibrated. It is shown that the system is capable of massive gas injections to provoke disruptions and delivering small puffs of gaseous impurities for perturbative transport experiments. This flexible tool is exploited in a series of impurity transport measurements with argon and neon injections. The impurities
First Results from Tests of High Temperature Superconductor Magnets on Tokamak
Energy Technology Data Exchange (ETDEWEB)
Gryaznevich, M.; Todd, T.T., E-mail: mikhail.gryaznevich@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Svoboda, V.; Markovic, T.; Ondrej, G. [Czech Technical University, Prague (Czech Republic); Stockel, J.; Duran, I.; Kovarik, K. [IPP Prague, Czech Technical University, Prague (Czech Republic); Sykes, A.; Kingham, D. [Tokamak Solutions, Culham Science Centre, Abingdon (United Kingdom); Melhem, Z.; Ball, S.; Chappell, S. [Oxford Instruments, Abingdon (United Kingdom); Lilley, M. K.; De Grouchy, P.; Kim, H. -T. [Imperial College, London (United Kingdom)
2012-09-15
construction of a small fully-HTS low aspect ratio tokamak has started at the Tokamak Solutions UK premises in the Culham Science Centre. It is planned to operate a small tokamak with A = 2 and circular cross section in steady state with plasma currents of 10 - 20 kA driven by Electron Bernstein Wave current drive. In parallel, the design and manufacture of a high-field (5 T) HTS TF coil for a spherical tokamak is carried out. (author)
Indian Academy of Sciences (India)
Our attraction to another body increases if the body is symmetricaland in proportion. If a face or a structure is in proportion,we are more likely to notice it and find it beautiful.The universal ratio of beauty is the 'Golden Ratio', found inmany structures. This ratio comes from Fibonacci numbers.In this article, we explore this ...
Indian Academy of Sciences (India)
Keywords. Fibonacci numbers, golden ratio, Sanskrit prosody, solar panel. Abstract. Our attraction to another body increases if the body is symmetricaland in proportion. If a face or a structure is in proportion,we are more likely to notice it and find it beautiful.The universal ratio of beauty is the 'Golden Ratio', found inmany ...
Indian Academy of Sciences (India)
Our attraction to another body increases if the body is sym- metrical and in proportion. If a face or a structure is in pro- portion, we are more likely to notice it and find it beautiful. The universal ratio of beauty is the 'Golden Ratio', found in many structures. This ratio comes from Fibonacci numbers. In this article, we explore this ...
MHD stability limits in the TCV Tokamak
International Nuclear Information System (INIS)
Reimerdes, H.
2001-07-01
Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation κ and triangularity δ, with high κ, and low δ leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The observed decrease
MHD stability limits in the TCV Tokamak
Energy Technology Data Exchange (ETDEWEB)
Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)
2001-07-01
Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The
Studies of non-inductive current drive in the CDX-U tokamak
International Nuclear Information System (INIS)
Hwang, Y.S.
1993-01-01
Two types of novel, non-inductive current drive concepts for starting-up and maintaining tokamak discharges, dc-helicity injection and internally-generated pressure-driven currents, have been developed on the CDX-U tokamak. To study the equilibrium and transport of these plasmas, a full set of magnetic diagnostics was installed. By applying a finite element method and a least squares error fitting technique, internal plasma current distributions are reconstructed from the measurements. Electron density distributions were obtained from 2 mm interferometer measurements by a similar least squares error technique utilizing magnetic flux configurations obtained by the magnetic analysis. Neoclassical pressure-driven currents in ECH plasmas are modeled with the reconstructed magnetic structure, using the electron density distribution and the electron temperature profile measured by a Langmuir probe. In the dc-helicity injection scheme, the need to increase injection current and maintain plasma equilibrium restricts possible arrangements. Several injection configurations were investigated, with the best found to be outside injection with a single divertor configuration, where the cathode is placed at the low field side of the x-point. Both pressure-driven and dc-helicity injected tokamaks show the importance of plasma equilibrium in obtaining high plasma current. Programmed vertical field operation has proven to be very important in achieving high plasma current. These non-inductive current drive techniques show great potential as efficient current drive methods for future steady-state and/or long-pulse fusion reactors
Cluster storage for COMPASS tokamak
Czech Academy of Sciences Publication Activity Database
Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír
2012-01-01
Roč. 87, č. 12 (2012), s. 2238-2241 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006
Surface tearing modes in tokamaks
International Nuclear Information System (INIS)
Takizuka, Tomonori; Kurita, Gen-ichi; Azumi, Masafumi; Takeda, Tatsuoki
1985-10-01
Surface tearing modes in tokamaks are studied numerically and analytically. The eigenvalue problem is solved to obtain the growth rate and the mode structure. We investigate in detail dependences of the growth rate of the m/n = 2/1 resistive MHD modes on the safety factor at the plasma surface, current profile, wall position, and resistivity. The surface tearing mode moves the plasma surface even when the wall is close to the surface. The stability diagram for these modes is presented. (author)
Major disruption process in tokamak
International Nuclear Information System (INIS)
Kurita, Gen-ichi; Azumi, Masafumi; Tuda, Takashi; Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji; Itoh, Kimitaka; Takeda, Tatsuoki
1981-11-01
The major disruption in a cylindrical tokamak is investigated by using the multi-helicity code, and the destabilization of the 3/2 mode by the mode coupling with the 2/1 mode is confirmed. The evolution of the magnetic field topology caused by the major disruption is studied in detail. The effect of the internal disruption on the 2/1 magnetic island width is also studied. The 2/1 magnetic island is not enhanced by the flattening of the q-profile due to the internal disruption. (author)
Simulations of drift waves in 3D magnetic configurations
International Nuclear Information System (INIS)
Jost, G.
2000-06-01
Drift waves are commonly held responsible for anomalous transport in tokamak configurations and in particular for the anomalously high heat loss. The next generation of stellarators on the other hand are hoped to be characterized by a much smaller neo-classical transport and by particle confinement close to that of tokamaks. There is nevertheless a strong interest in the stellarator community to study the properties of drift waves in 3D magnetic configurations. To serve this interest we have developed the first global gyrokinetic code, EUTERPE, aimed at the investigation of linear drift wave stability in general toroidal geometry. The physical model assumes electrostatic waves and adiabatic electrons. EUTERPE is a particle-in-cell (PIC) code in which the gyrokinetic Poisson equation is discretized with the finite element method defined in the PEST -1 system of magnetic coordinates. The magnetic geometry is provided by the magnetohydrodynamic (MHD) equilibrium code VMEC. The complete 3D model has been successfully validated in toroidal axisymmetric and straight helical geometries and has permitted the first simulation of unstable global ITG driven modes in non-axisymmetric toroidal configurations. As a first application, two configurations have been studied, the Quasi-Axially symmetric Stellarator with three fields periods (QAS3) currently one system under consideration at the Princeton Plasma Physics Laboratory and the Helically Symmetric experiment (HSX) which has recently started operation at the University of Wisconsin. QAS3 is characterized by a tokamak-Iike field in the outer part of the torus. In this structure the drift waves are mainly affected by the magnetic shear and barely by the shape of the plasma. Also, the results are very close to those obtained for a tokamak. On the other hand, results for the HSX configuration, which is characterized by a dominant helical magnetic field, show a clear 3D effect, namely a strong toroidal variation of the drift wave
VENUS+δf - A bootstrap current calculation module for 3D configurations
International Nuclear Information System (INIS)
Isaev, M.Yu.; Brunner, S.; Cooper, W.A.; Tran, T.M.; Bergmann, A.; Beidler, C.D.; Geiger, J.; Maassberg, H.; Nuehrenberg, J.; Schmidt, M.
2005-01-01
We present a new 3D code VENUS+δf for neoclassical transport calculations in nonaxisymmetric toroidal systems. Numerical drift orbits from the original VENUS code and the δf method for tokamak transport calculations are combined. The first results obtained with VENUS+δf are compared with neoclassical theory for different collisional regimes in a JT-60 tokamak test case with monoenergetic particles and with a Maxwellian distribution. Benchmarks with DKES code results for the bootstrap current in the W7X configuration as well as further VENUS+δf developments are discussed. (author)
Simulator configuration maintenance
International Nuclear Information System (INIS)
2006-01-01
Requirements and recommendations of this section defines NPP personnel activity aimed to the provision of the simulator configuration compliance with the current configuration of the power-generating unit-prototype, standard and technical requirements and describe a monitoring procedure for a set of simulator software and hardware, training, organizational and technical documents
PIV Logon Configuration Guidance
Energy Technology Data Exchange (ETDEWEB)
Lee, Glen Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-03-04
This document details the configurations and enhancements implemented to support the usage of federal Personal Identity Verification (PIV) Card for logon on unclassified networks. The guidance is a reference implementation of the configurations and enhancements deployed at the Los Alamos National Laboratory (LANL) by Network and Infrastructure Engineering – Core Services (NIE-CS).
Business Model Process Configurations
DEFF Research Database (Denmark)
Taran, Yariv; Nielsen, Christian; Thomsen, Peter
2015-01-01
, by developing (inductively) an ontological classification framework, in view of the BM process configurations typology developed. Design/methodology/approach – Given the inconsistencies found in the business model studies (e.g. definitions, configurations, classifications) we adopted the analytical induction...
Study of intelligent system for control of the tokamak-ETE plasma positioning
International Nuclear Information System (INIS)
Barbosa, Luis Filipe de Faria Pereira Wiltgen
2003-01-01
The development of an intelligent neural control system of the neural type, capable to perform real time control of the plasma displacement in the experiment tokamak spheric - ETE (spherical tokamak experiment ) is presented. The ETE machine is in operation since Nov 2000, in the LAP - Plasma Associated Laboratory of the Brazilian Institute on Spatial Research (INPE) in Sao Jose dos Campos, S P, Brazil. The experiment is dedicated to study the magnetic confinement of a fusion plasma in a configuration favorable for the construction of future reactors. Nuclear fusion constitutes a renewable energy source with low environmental impact, which uses atomic energy in pacific applications for the sustainable development of humanity. One of the important questions for the attainment of fusion relates to the stability of the plasma and control of its position during the reactor operation. Therefore, the development of systems to control the plasma in tokamaks constitutes a necessary technological advance for the feasibility of nuclear fusion. In particular, the research carried out in this thesis concerns the proposal of a system to control the vertical displacement of the plasma in the ETE tokamak, aiming to obtain steady pulses in this machine. A Magnetic Levitation system (Mag Lev) was developed as part of this work, allowing to study the nonlinear behavior of a device that, from the aspect of position control, is similar (analogous) to the plasma in the ETE tokamak, This magnetic levitation system was designed, mathematically modeled and built in order to test both classical and intelligent type controllers. The results of this comparison are very promising for the use of intelligent controllers in the ETE tokamak as well as other control applications. (author)
The benefits of ITER for the portfolio of fusion configurations
International Nuclear Information System (INIS)
Goldston, R.J.
2002-01-01
Recent plasma science challenges are 1) what limits the pressure in plasmas? (macroscopic stability), 2) how do hot particles and plasma waves interact in the non-linear regime? (wave-particle interactions), 3) what causes plasma transport? (microscopic turbulence and transport) and 4) how can high-temperature plasma and material surface co-exist? (plasma-material interactions). This fusion plasma science is addressed using a 'Portfolio' of configurations, like Stellarator, Tokamak, Spherical Torus, Reversed Field Pinch, Spheromak, and Field Reversed Configuration. Namely, the scientific results from one configuration benefit progress in others. Recent example of this effort can be found in NCSX, NSTX and RFP. ITER will provide very significant benefits to the development of the full fusion portfolio; macroscopic stability, wave-particle interactions, microturbulence and transport, plasma-material interactions, and technical demonstration of an integrated fusion system. (author)
The benefits of ITER for the portfolio of fusion configurations
Energy Technology Data Exchange (ETDEWEB)
Goldston, R.J. [Princeton Plasma Physics Lab., NJ (United States)
2002-10-01
Recent plasma science challenges are 1) what limits the pressure in plasmas? (macroscopic stability), 2) how do hot particles and plasma waves interact in the non-linear regime? (wave-particle interactions), 3) what causes plasma transport? (microscopic turbulence and transport) and 4) how can high-temperature plasma and material surface co-exist? (plasma-material interactions). This fusion plasma science is addressed using a 'Portfolio' of configurations, like Stellarator, Tokamak, Spherical Torus, Reversed Field Pinch, Spheromak, and Field Reversed Configuration. Namely, the scientific results from one configuration benefit progress in others. Recent example of this effort can be found in NCSX, NSTX and RFP. ITER will provide very significant benefits to the development of the full fusion portfolio; macroscopic stability, wave-particle interactions, microturbulence and transport, plasma-material interactions, and technical demonstration of an integrated fusion system. (author)
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)
Risk-based configuration control
International Nuclear Information System (INIS)
Szikszai, T.
1997-01-01
The presentation discusses the following issues: The Configuration Control; The Risk-based Configuration Control (during power operation mode, and during shutdown mode). PSA requirements. Use of Risk-based Configuration Control System. Configuration Management (basic elements, benefits, information requirements)