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Sample records for spherical tokamak development

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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)

  7. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  8. 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)

  9. 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

  10. 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.

  11. Compact fusion energy based on the spherical tokamak

    Science.gov (United States)

    Sykes, A.; Costley, A. E.; Windsor, C. G.; Asunta, O.; Brittles, G.; Buxton, P.; Chuyanov, V.; Connor, J. W.; Gryaznevich, M. P.; Huang, B.; Hugill, J.; Kukushkin, A.; Kingham, D.; Langtry, A. V.; McNamara, S.; Morgan, J. G.; Noonan, P.; Ross, J. S. H.; Shevchenko, V.; Slade, R.; Smith, G.

    2018-01-01

    Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power using relatively small devices. We present an overview of the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

  12. 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.)

  13. Alfven Eigenmodes in spherical tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, Mikhail P.; Sharapov, Sergei E.; Berk, Herbert L.; Pinches, Simon D.

    2005-01-01

    Electromagnetic instabilities are often excited by fast super-Alfvenic ions produced by neutral beam injection (NBI) in plasmas of the spherical tokamaks START and MAST (toroidal magnetic confinement devices in which the minor a and major R 0 radii of the torus are comparable, R 0 /a≅1.2/1.8). These instabilities are seen as discrete weakly-damped toroidal and elliptical Alfven Eigenmodes (TAEs and EAEs) with frequencies tracing in time the Alfven scaling with the equilibrium magnetic field and plasma density, or as energetic particle modes (EPMs) whose frequencies don't start from TAE-frequency and sweep down in time faster than the equilibrium parameters change. In some discharges the beam drives Aflvenic-type modes that start from the TAE frequency and sweep in both up- and down- directions. Such electromagnetic perturbations are interpreted as 'hole-clump' long-living nonlinear fluctuations of the fast ion distribution function predicted by Berk-Breizman-Petviashvili [Phys. Lett. A238 (1998) 408]. It is found on both START and MAST that the Alfven instabilities weaken in their mode amplitude and in the number of unstable modes as the pressure of the thermal plasma increases, in agreement with increased thermal ion Landau damping and the pressure effect on core-localised TAEs. (author)

  14. Spherical tokamak without external toroidal fields

    International Nuclear Information System (INIS)

    Kaw, P.K.; Avinash, K.; Srinivasan, R.

    2001-01-01

    A spherical tokamak design without external toroidal field coils is proposed. The tokamak is surrounded by a spheromak shell carrying requisite force free currents to produce the toroidal field in the core. Such equilibria are constructed and it is indicated that these equilibria are likely to have robust ideal and resistive stability. The advantage of this scheme in terms of a reduced ohmic dissipation is pointed out. (author)

  15. Spherical tokamak power plant design issues

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  16. 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)

  17. Fusion technology applications of the spherical tokamak

    International Nuclear Information System (INIS)

    Robinson, D.C.; Akers, R.; Allfrey, S.J.

    1999-01-01

    Fusion technology applications of the spherical tokamak are presented, exploiting its high β capability, normal conducting TF coils, compact core, high natural elongation, disruption resilience and low capital cost. We concentrate here on two particular applications: a volume neutron source (VNS) for component testing and a power plant, addressing engineering and physics issues for steady state operation. The prospect of nearer term burning plasma ST devices are discussed in the conclusions. (author)

  18. Fusion technology applications of the spherical tokamak

    International Nuclear Information System (INIS)

    Robinson, D.C.; Akers, R.; Allfrey, S.J.

    2001-01-01

    Fusion technology applications of the spherical tokamak are presented, exploiting its high β capability, normal conducting TF coils, compact core, high natural elongation, disruption resilience and low capital cost. We concentrate here on two particular applications: a volume neutron source (VNS) for component testing and a power plant, addressing engineering and physics issues for steady state operation. The prospect of nearer term burning plasma ST devices are discussed in the conclusions. (author)

  19. The spherical tokamak fusion power plant

    International Nuclear Information System (INIS)

    Wilson, H.R.; Voss, G.; Ahn, J.W.

    2003-01-01

    The design of a 1GW(e) steady state fusion power plant, based on the spherical tokamak concept, has been further iterated towards a fully self-consistent solution taking account of plasma physics, engineering and neutronics constraints. In particular a plausible solution to exhaust handling is proposed and the steam cycle refined to further improve efficiency. The physics design takes full account of confinement, MHD stability and steady state current drive. It is proposed that such a design may offer a fusion power plant which is easy to maintain: an attractive feature for the power plants following ITER. (author)

  20. 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)

  1. Application studies of spherical tokamak plasma merging

    International Nuclear Information System (INIS)

    Ono, Yasushi; Inomoto, Michiaki

    2012-01-01

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

  2. Excitation of Alfvenic instabilities in spherical tokamaks

    International Nuclear Information System (INIS)

    McClements, K.G.; Appel, L.C.; Hole, M.J.; Thyagaraja, A.

    2003-01-01

    Understanding energetic particle confinement in spherical tokamak (STs) is important for optimising the design of ST power plants, and provides a testbed for theoretical modelling under conditions of strong toroidicity and shaping, and high beta. MHD analysis of some recent beam-heated discharges in the MAST ST indicates that high frequency modes observed in these discharges can be identified as toroidal Alfven Eigenmodes (TAEs) and elliptical Alfven Eigenmodes (EAEs). It is possible that such modes could strongly enhance fusion alpha-particle transport in an ST power plant. Computations of TAE growth rates for one particular MAST discharge, made using the HAGIS guiding centre code and benchmarked against analytical estimates, indicate strong drive by sub-Alfvenic neutral beam ions. HAGIS computations using higher mode amplitudes than those observed indicate that whereas co-passing beam ions provide the bulk of he TAE drive, counter-passing ions provide the dominant component of TAE-induced particle losses. Axisymmetric Alfvenic mode activity has been detected during ohmic discharges in MAST. These observations are shown by computational modelling to be consistent with the excitation of global Alfven Eigenmodes (GAEs) with n=0 and low m, driven impulsively by low frequency MHD. (author)

  3. 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

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

    International Nuclear Information System (INIS)

    Isayama, Akihiko; Tanaka, Hitoshi

    2009-01-01

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

  5. Merging startup experiments on the UTST spherical tokamak

    International Nuclear Information System (INIS)

    Yamada, Takuma; Kamio, Shuji; Imazawa, Ryota

    2010-01-01

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

  6. Fishbone mode in high-β discharges of spherical tokamaks

    International Nuclear Information System (INIS)

    Kolesnichenko, Ya.I.; Lutsenko, V.V.; Marchenko, V.S.

    2000-01-01

    Using Hamiltonian formalism, it has been shown that well-trapped energetic ions moving outwards consume the energy of MHD perturbations through the precessional resonance provided that the plasma pressure is sufficiently high. This supports the conclusion of recent publication that the fishbone mode is stabilized in high-β discharges of spherical tokamaks. It has also been found that the presence of the velocity anisotropy of energetic ions does not change this conclusion. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  8. The Flinders University inductively driven spherical Tokamak project

    International Nuclear Information System (INIS)

    McCarthy, L.

    1998-01-01

    Full text: The Flinders University inductive start up Spherical Tokamak (ST) program is designed with two major functions: first a target plasma for a definitive test of rotating magnetic field (RMF) current drive, and secondly as a target plasma to be used in development of diagnostics for the collaboration between Flinders University and the Australian National Fusion Facility. A third goal is to maintain an Australian link to the international ST community at a time when this ST approach to plasma fusion is entering a ''second generation'' phase of larger machines, following the demonstration of resilience to major disruptions on START and MEDUSA, and excellent confinement properties, and β. Modelling of the optimum operating regime consistent with power supplies available at Flinders University, and comparisons of plasmas prepared by RMF alone with ohmically heated plasmas such as START, are presented to support the need for the design of this OH hot confined target plasma approach to RMF current drive as an alternative to that of pure RMF current drive at higher powers being attempted elsewhere, should that approach not prove successful. Progress on the experiments, which now includes successful tests of the toroidal field system and the OH coil system, is reported. The RMF facility will not be available till late in 1998. The case is made for retaining the valuable equipment resources of the Flinders University plasma research group and negotiating for the transfer of these to the Australian National Fusion Facility at the completion of this project at the end of 1999

  9. Present status of operation of the ETE spherical tokamak

    International Nuclear Information System (INIS)

    Bosco, E. del; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Ludwig, G.O.; Shibata, C.S.

    2005-01-01

    The ETE is a spherical tokamak with aspect ratio A = 1.5 (major radius of 0.3m and minor radius of 0.2m) under development at LAP/INPE. The ETE incorporates some innovative features that resulted in a compact and light weighted device with good plasma accessibility. Since the first plasma obtained at the very end of 2000 (Ip = 12kA, duration of 2ms, B o = 0.1T), the machine is operational and improvements are being done in order to achieve the planned final parameter values for the first phase of operation (Ip = 220kA, duration 15ms, B o = 0.4T), which are limited by the available capacitors. The efforts are being focused on incrementing the energy of the capacitor banks, lessening the stray magnetic fields in the plasma region, conditioning the vacuum vessel wall, implementing diagnostics and optimizing the discharge parameters. Presently, plasma currents in the range of 40-60kA (duration of 6-12 ms) are routinely obtained. Electron temperatures up to 160eV and plasma densities up to 3.0x10 19 m -3 are being reached. (author)

  10. Start-up of spherical tokamak without a center solenoid

    International Nuclear Information System (INIS)

    Maekawa, Takashi; Nagata, Masayoshi

    2012-01-01

    For low-aspect tokamak reactors, spherical tokamak reactors, ST-type FESF/CTFs, it is essential to remove or minimize a central solenoid (CS). Even with the minimized CS, non-inductive start up of the plasma current is required. Rapid increase in the spontaneous plasma current at the final stage of current start-up drives ignition. At the initial stage, formation of plasma and magnetic surfaces are required. As non-inductive plasma start-up scenarios, ECH/ECCD, LHCD, HHFW, DC HELICITY injection, plasma merging and NBI have been studied. In the present article, the present status and future prospect of experimental and theoretical works on these subjects. (author)

  11. New results from the Globus-M spherical tokamak

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  12. New results from Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.

    2002-01-01

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

  13. Numerical simulation of internal reconnection event in spherical tokamak

    International Nuclear Information System (INIS)

    Hayashi, Takaya; Mizuguchi, Naoki; Sato, Tetsuya

    1999-07-01

    Three-dimensional magnetohydrodynamic simulations are executed in a full toroidal geometry to clarify the physical mechanisms of the Internal Reconnection Event (IRE), which is observed in the spherical tokamak experiments. The simulation results reproduce several main properties of IRE. Comparison between the numerical results and experimental observation indicates fairly good agreements regarding nonlinear behavior, such as appearance of localized helical distortion, appearance of characteristic conical shape in the pressure profile during thermal quench, and subsequent appearance of the m=2/n=1 type helical distortion of the torus. (author)

  14. The prospects for electron Bernstein wave heating of spherical tokamaks

    International Nuclear Information System (INIS)

    Cairns, R.A.; Lashmore-Davies, C.N.

    2000-02-01

    Electron Bernstein waves are analysed as possible candidates for heating spherical tokamaks. An inhomogeneous plane slab model of the plasma with a sheared magnetic field is used to calculate the linear conversion of the ordinary mode (O-mode) to the extraordinary mode (X-mode). A formula for the fraction of the incident O-mode energy which is converted to the X-mode at the O-mode cut-off is derived. This fraction is then able to propagate to the upper hybrid resonance where it is converted to the electron Bernstein mode. The damping of electron Bernstein waves at the fourth harmonic resonance, corresponding to a 60GHz source on the Mega Amp Spherical Tokamak MAST [A C Darke et al Proc 16th Symposium on Fusion Energy, Champaign- Urbana, Illinois USA IEEE, 2 p1456 (1995)], is computed. This is shown to be so strongly absorbing that the electron Bernstein wave would be totally absorbed in the outer regions of the resonance. This feature implies that electron Bernstein wave current drive (on- or off-axis) could be very efficient. (author)

  15. 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)

  16. Neutronics design for a spherical tokamak fusion-transmutation reactor

    International Nuclear Information System (INIS)

    Deng Meigen; Feng Kaiming; Yang Bangchao

    2002-01-01

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

  17. Overview and initial results of the ETE spherical tokamak

    International Nuclear Information System (INIS)

    Berni, L.A.; Del Bosco, E.; Ferreira, J.G.; Ludwig, G.O.; Oliveira, R.M.; Shibata, C.S.; Barbosa, L.F.F.P.W.; Vilela, W.A.

    2003-01-01

    The ETE spherical tokamak is a small size aspect-ratio machine with major and minor radius of 30 cm and 20 cm, respectively. The vessel was made of Inconel 625 and provides good access for plasma diagnostics through 58 Conflat ports. The first plasma was obtained at the end of 2000 and presently plasma currents of about 45 kA lasting for about 4 ms with electron temperature up to 160 eV and densities of 2.2x10 19 m -3 are routinely obtained. Achievement of the designed parameters for the first phase of operation is expected by the end of this year, with plasma current up to 200 kA lasting for about 15 ms. This paper describes some details of the ETE project, construction and mainly the first results and analysis of basic parameters. (author)

  18. 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)

  19. Stability at high performance in the MAST spherical tokamak

    International Nuclear Information System (INIS)

    Buttery, R.J.; Akers, R.; Arends, E. =

    2003-01-01

    The development of reliable H-modes on MAST, together with advances in heating power and a range of powerful diagnostics, has provided a platform to enable MAST to address some of he most important issues of tokamak stability. In particular the high β potential of the ST is highlighted with stable operation at β N ∼5-6 , β T ∼ 16% and β p as high as 1.9, confirmed by a range of profile diagnostics. Calculations indicate that β N levels are in the vicinity of no-wall stability limits. Studies have provided the first identification of the Neoclassical Tearing Mode (NTM) in the ST, using its behaviour to quantitatively validate predictions of NTM theory, previously only applied to conventional tokamaks. Experiments have demonstrated that sawteeth play a strong role in triggering NTMs - by avoiding large sawteeth much higher β N can, and has, been reached. Further studies have confirmed the NTM's significance, with large islands observed using the 300 point Thomson diagnostic, and locking of large n=1 modes frequently leading to disruptions. H-mode plasmas are also limited by ELMs, with confinement degraded as ELM frequency rises. However, unlike the conventional tokamak, the ELMs in high performing regimes on MAST (H IPB98Y2 ∼1) appear to be type III in nature. Modelling identifies instability to peeling modes, consistent with a type III interpretation, and shows considerable scope to raise pressure gradients (despite n=∞ ballooning theory predictions of instability) before ballooning type modes (perhaps associated with type I ELMs) occur. Finally sawteeth are shown not to remove the q=1 surface in the ST - other promising models are being explored. Thus research on MAST is not only demonstrating stable operation at high performance levels, and developing methods to control instabilities; it is also providing detailed tests of the stability physics and models applicable to conventional tokamaks, such as ITER. (author)

  20. Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR

    Science.gov (United States)

    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.

  1. Formation of transport barriers in the MAST spherical tokamak

    International Nuclear Information System (INIS)

    Meyer, H; Field, A R; Akers, R J; Brickley, C; Conway, N J; Patel, A; Carolan, P G; Challis, C; Counsell, G F; Cunningham, G; Helander, P; Kirk, A; Lloyd, B; Maingi, R; Tournianski, M R; Walsh, M J

    2004-01-01

    In the Mega Ampere Spherical Tokamak (MAST) plasmas have been generated with internal (ITB) or edge (ETB) transport barriers. ITBs were achieved in both the electron and the ion energy channel. In the presence of an ITB in the ion energy channel, transport analysis shows that the ion thermal diffusivity, χ i , is reduced to almost neoclassical values while the ITB persists. The widely tested criteria for ITB formation ρ t * =ρ s αlnT/αR>ρ ITB * ∼0.014 (ρ s : Larmor radius at sound speed) obtained from dimensional analysis of JET discharges is easily exceeded on MAST. Even without the evidence of an ρ T * >0.014 often applies, showing that this criterion in its current form is not generally applicable. ETBs are most easily formed in MAST if in a double null divertor configuration the discharge is vertically balanced, so that both X-points are almost on the same flux surface (CDND), and if the plasma is refuelled from the high field side mid-plane. The H-mode threshold power, P thr = 0.5 MW, in connected double null diverted (CDND) is only about half of that in a similar disconnected discharge with the ion ∇ B drift towards the X-point on the last closed flux surface (LDND). P thr scales between lower double null diverted (LDND) and the single null diverted configuration with the plasma surface area on MAST

  2. Study on wall recycling behaviour in CPD spherical tokamak

    International Nuclear Information System (INIS)

    Bhattacharyay, R.; Zushi, H.; Hirooka, Y.; Sakamoto, M.; Yoshinaga, T.; Okamoto, K.; Kawasaki, S.; Hanada, K.; Sato, K.N.; Nakamura, K.; Idei, H.; Ryoukai, T.; Nakashima, H.; Higashijima, A.

    2008-01-01

    Experiments to study wall recycling behaviour have been performed in the small spherical tokamak compact plasma-wall interaction experimental device (CPD) from the viewpoint of global as well as local plasma wall interaction condition. Electron cyclotron resonance (ECR) plasma of typically ∼50 to 400 ms duration is produced using ∼40 to 80 kW RF power. In order to study the global wall recycling behaviour, pressure measurements are carried out just before and after the ECR plasma in the absence of any external pumping. The recycling behaviour is found to change from release to pumping beyond a certain level of pressure value which is again found to be a function of shot history. The real-time local wall behaviour is studied in similar RF plasma using a rotating tungsten limiter, actively coated with lithium. Measurement of H α light intensity in front of the rotating surface has indicated a clear reduction (∼10%) in the steady-state hydrogen recycling with continuous Li gettering of several minutes

  3. Nonlinear simulation of edge-localized mode in spherical tokamak

    International Nuclear Information System (INIS)

    Mizuguchi, N.; Hayashi, T.; Nakajima, N.; Khan, R.

    2006-10-01

    A numerical modeling for the dynamics of an edge-localized mode (ELM) crash in the spherical tokamak is proposed with a consecutive scenario which is initiated by the spontaneous growth of the ballooning mode instability by means of a three-dimensional nonlinear magnetohydrodynamic simulation. The simulation result shows a two-step relaxation process which is induced by the intermediate-n ballooning instability followed by the m/n=1/1 internal kink mode, where m and n represent the poloidal and toroidal mode numbers, respectively. By comparing with the experimental observations, we have found that the simulation result can reproduce several characteristic features of the so-called type-I ELM in an appropriate time scale: (1) relation to the ballooning instability, (2) intermediate-n precursors, (3) low-n structure on the crash, (4) formation and separation of the filament, and (5) considerable amount of loss of plasma. Furthermore, the model is verified by examining the effect of diamagnetic stabilization and comparing the nonlinear behavior with that of the peeling modes. The ion diamagnetic drift terms are found to stabilize some specific components linearly; nevertheless they are not so effective in the nonlinear dynamics such as the filament formation and the amount of loss. For the peeling mode case, no prominent filament structure is formed in contrast to the ballooning case. (author)

  4. Physics objectives of PI3 spherical tokamak program

    Science.gov (United States)

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

    2017-10-01

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

  5. Plasma rotation and transport in MAST spherical tokamak

    Science.gov (United States)

    Field, A. R.; Michael, C.; Akers, R. J.; Candy, J.; Colyer, G.; Guttenfelder, W.; Ghim, Y.-c.; Roach, C. M.; Saarelma, S.; MAST Team

    2011-06-01

    The formation of internal transport barriers (ITBs) is investigated in MAST spherical tokamak plasmas. The relative importance of equilibrium flow shear and magnetic shear in their formation and evolution is investigated using data from high-resolution kinetic- and q-profile diagnostics. In L-mode plasmas, with co-current directed NBI heating, ITBs in the momentum and ion thermal channels form in the negative shear region just inside qmin. In the ITB region the anomalous ion thermal transport is suppressed, with ion thermal transport close to the neo-classical level, although the electron transport remains anomalous. Linear stability analysis with the gyro-kinetic code GS2 shows that all electrostatic micro-instabilities are stable in the negative magnetic shear region in the core, both with and without flow shear. Outside the ITB, in the region of positive magnetic shear and relatively weak flow shear, electrostatic micro-instabilities become unstable over a wide range of wave numbers. Flow shear reduces the linear growth rates of low-k modes but suppression of ITG modes is incomplete, which is consistent with the observed anomalous ion transport in this region; however, flow shear has little impact on growth rates of high-k, electron-scale modes. With counter-NBI ITBs of greater radial extent form outside qmin due to the broader profile of E × B flow shear produced by the greater prompt fast-ion loss torque.

  6. Initial results from the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Takase, Y.; Ejiri, A.; Kasuya, N.

    2001-01-01

    A new spherical tokamak TST-2 was constructed at the University of Tokyo and started operation in September 1999. Reliable plasma initiation is achieved with typically 1 kW of ECH power at 2.45 GHz. Plasma currents of up to 90 kA and toroidal fields of up to 0.2 T have been achieved during the initial experimental campaign. The ion temperature is typically 100 eV. Internal reconnection events (IREs) are often observed. The internal magnetic field measured at r/a=2/3 indicated growth of fluctuations up to the 4 th harmonic, suggesting the existence of modes with several different mode numbers. In the presence of a toroidal field and a vertically oriented mirror field, noninductively driven currents of order 1 kA were observed with 1 kW of ECH power. The driven current increased with decreasing filling pressure, down to 3x10 -6 torr. A study of high harmonic fast wave (HHFW) excitation and propagation has begun. Initial results indicate highly efficient wave launching. (author)

  7. EBW H&CD Potential for Spherical Tokamaks

    Science.gov (United States)

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

    2011-12-01

    Spherical tokamaks (STs), which feature relatively high neutron flux and good economy, operate generally in high-ß regimes, in which the usual EC O- and X- modes are cut-off. In this case, electron Bernstein waves (EBWs) seem to be the only option that can provide features similar to the EC waves—controllable localized heating and current drive (H&) that can be utilized for core plasma heating as well as for accurate plasma stabilization. We first derive an analytical expression for Gaussian beam OXB conversion efficiency. Then, an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX) is performed. Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

  8. 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.

  9. 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

  10. Formation of transport barriers in the MAST spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, H [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Field, A R [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Akers, R J [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Brickley, C [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Conway, N J [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Patel, A [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Carolan, P G [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Challis, C [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Counsell, G F [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Cunningham, G [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Helander, P [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Kirk, A [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Lloyd, B [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Maingi, R [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Tournianski, M R [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Walsh, M J [Walsh Scientific Ltd, Culham Science Centre, Abingdon, Oxfordshire, OX14 3EB (United Kingdom)

    2004-05-01

    In the Mega Ampere Spherical Tokamak (MAST) plasmas have been generated with internal (ITB) or edge (ETB) transport barriers. ITBs were achieved in both the electron and the ion energy channel. In the presence of an ITB in the ion energy channel, transport analysis shows that the ion thermal diffusivity, {chi}{sub i}, is reduced to almost neoclassical values while the ITB persists. The widely tested criteria for ITB formation {rho}{sub t}{sup *}={rho}{sub s}{alpha}lnT/{alpha}R>{rho}{sub ITB}{sup *}{approx}0.014 ({rho}{sub s}: Larmor radius at sound speed) obtained from dimensional analysis of JET discharges is easily exceeded on MAST. Even without the evidence of an {rho}{sub T}{sup *}>0.014 often applies, showing that this criterion in its current form is not generally applicable. ETBs are most easily formed in MAST if in a double null divertor configuration the discharge is vertically balanced, so that both X-points are almost on the same flux surface (CDND), and if the plasma is refuelled from the high field side mid-plane. The H-mode threshold power, P{sub thr} = 0.5 MW, in connected double null diverted (CDND) is only about half of that in a similar disconnected discharge with the ion {nabla} B drift towards the X-point on the last closed flux surface (LDND). P{sub thr} scales between lower double null diverted (LDND) and the single null diverted configuration with the plasma surface area on MAST.

  11. The National Spherical Tokamak Experiment at the Princeton Plasma Physics Laboratory

    International Nuclear Information System (INIS)

    1995-12-01

    The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1108, evaluating the environmental effects of the proposed construction and operation of the National Spherical Tokamak Experiment (NSTX) within the existing C-Stellarator (CS) Building at the Princeton Plasma Physics Laboratory, Princeton, New Jersey. The purpose of the NSTX is to investigate the physics of spherically shaped plasmas as an alternative path to conventional tokamaks for development of fusion energy. Fusion energy has the potential to help compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Construction of the NSTX in the CS Building would require the dismantling and removal of the existing unused Princeton Large Torus (PLT) device, part of which would be reused to construct the NSTX. Based on the analyses in the EA, the DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 U.S.C. 4,321 et seq. The preparation of an Environmental Impact Statement is not required. Thus, the DOE is issuing a FONSI pursuant to the Council on Environmental Quality regulations implementing NEPA (40 CFR Parts 1500--1508) and the DOE NEPA implementing regulations (10 CFR Part 1021)

  12. Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR

    Science.gov (United States)

    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.

  13. 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)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  15. L-mode SOL width scaling in the MAST spherical tokamak

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  16. L-H transition in the mega-Amp spherical tokamak

    DEFF Research Database (Denmark)

    Akers, R.J.; Counsell, G.F.; Sykes, A.

    2002-01-01

    H-mode plasmas have been achieved on the MAST spherical tokamak at input power considerably higher than predicted by conventional threshold scalings. Following L-H transition, a clear improvement in energy confinement is obtained, exceeding recent international scalings even at densities approach...

  17. First physics results from the MAST Mega-Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Sykes, A.; Ahn, J.-W.; Akers, R.; Arends, E.; Carolan, P.G.; Counsell, G.F.; Fielding, S. J.; Gryaznevich, M.; Martin, R.; Price, M.; Roach, C.; Shevchenko, V.; Tournianski, M.; Valovic, M.; Walsh, M.J.; Wilson, H.R.

    2001-01-01

    First physics results are presented from MAST (Mega-Amp Spherical Tokamak), one of the new generation of purpose built spherical tokamaks (STs) now commencing operation. Some of these results demonstrate, for the first time, the novel effects of low aspect ratio, for example, the enhancement of resistivity due to neo-classical effects. H-mode is achieved and the transition to H-mode is accompanied by a tenfold steepening of the edge density gradient which may enable the successful application of electron Bernstein wave heating in STs. Studies of halo currents show that these less than expected from conventional tokamak results, and measurements of divertor power loading confirm that most of the power flows to the outer strike points, easing the power handling on the inner points (a critical issue for STs)

  18. High kinetic energy plasma jet generation and its injection into the Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Voronin, A.V.; Gusev, V.K.; Petrov, Yu.V.; Sakharov, N.V.; Abramova, K.B.; Sklyarova, E.M.; Tolstyakov, S.Yu.

    2005-01-01

    Progress in the theoretical and experimental development of the plasma jet source and injection of hydrogen plasma and neutral gas jets into the Globus-M spherical tokamak is discussed. An experimental test bed is described for investigation of intense plasma jets that are generated by a double-stage plasma gun consisting of an intense source for neutral gas production and a conventional pulsed coaxial accelerator. A procedure for optimizing the accelerator parameters so as to achieve the maximum possible flow velocity with a limited discharge current and a reasonable length of the coaxial electrodes is presented. The calculations are compared with experiment. Plasma jet parameters, among them pressure distribution across the jet, flow velocity, plasma density, etc, were measured. Plasma jets with densities of up to 10 22 m -3 , total numbers of accelerated particles (1-5) x 10 19 , and flow velocities of 50-100 km s -1 were successfully injected into the plasma column of the Globus-M tokamak. Interferometric and Thomson scattering measurements confirmed deep jet penetration and a fast density rise ( 19 to 1 x 10 19 ) did not result in plasma degradation

  19. Experimental study on practicability of self-created spherical tokamak in coilless STPC-EX machine

    International Nuclear Information System (INIS)

    Sinman, S.

    2002-01-01

    The aim of this study is to recognize the physical basis of the alternative self organization mechanism occurred STPC-EX machine. The conventional diagnostic tools are used in this study and for photographic recording, open shutter integrated post-fogging method is preferred. The annular coaxial two plasma current sheets one within other at the same direction are created and flowed on the surface of floating conductive central rod. Consequently, spherical tokamak configurated by new creation mechanism of Dual Axial Z-Pinch. (DAZP) yields fairly high beta of 0.4-0.6 at self created spherical tokamak plasma. Sustainment time of DAZP is 5.6-6.3 mili second. (author)

  20. Sustainment of spherical tokamak by means of repetitive injection of compact torus plasma

    International Nuclear Information System (INIS)

    Shimamura, Shin; Matsura, Ken; Takahashi, Tsutomu; Nogi, Yasuyuki

    2000-01-01

    Sustainment of spherical tokamak (S.T.) has been studied. A compact torus (C.T.) plasma was injected into confinement region by magnetized coaxial gun. For start-up and sustainment of large main spherical tokamak, single pulsed injection of small C.T. is not sufficient in many cases. C.T.plasma injection of high repetition rate is required. For this purpose magnetized coaxial gun was driven with high repetition rate current. The first injected C.T. plasma could start-up S.T. without other help. The repetitive C.T. injection grew and sustained the S.T. plasma. A CCD camera with fast gated image intensifier took a cross sectional view of S.T. during the repetitive C.T. injection. (author)

  1. Bounce Precession Fishbones in the National Spherical Tokamak Experiment

    International Nuclear Information System (INIS)

    Eric Fredrickson; Liu Chen; Roscoe White Eric Fredrickson; Liu Chen; Roscoe White

    2003-01-01

    Bursting modes are observed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40 (2000) 557], which are identified as bounce-precession-frequency fishbone modes. They are predicted to be important in high-current, low-shear discharges with a significant population of trapped particles with a large mean-bounce angle, such as produced by near-tangential beam injection into a large aspect-ratio device. Such a distribution is often stable to the usual precession-resonance fishbone mode. These modes could be important in ignited plasmas, driven by the trapped-alpha-particle population

  2. A conceptual design of superconducting spherical tokamak reactor

    International Nuclear Information System (INIS)

    Nagayama, Yoshio; Shinya, Kichiro; Tanaka, Yasutoshi

    2012-01-01

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

  3. 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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  6. Nonlocal neoclassical transport in tokamak and spherical torus experiments

    International Nuclear Information System (INIS)

    Wang, W. X.; Rewoldt, G.; Tang, W. M.; Hinton, F. L.; Manickam, J.; Zakharov, L. E.; White, R. B.; Kaye, S.

    2006-01-01

    Large ion orbits can produce nonlocal neoclassical effects on ion heat transport, the ambipolar radial electric field, and the bootstrap current in realistic toroidal plasmas. Using a global δf particle simulation, it is found that the conventional local, linear gradient-flux relation is broken for the ion thermal transport near the magnetic axis. With regard to the transport level, it is found that details of the ion temperature profile determine whether the transport is higher or lower when compared with the predictions of standard neoclassical theory. Particularly, this nonlocal feature is suggested to exist in the National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, Y.-K. M. Peng et al., Nucl. Fusion 40, 557 (2000)], being consistent with NSTX experimental evidence. It is also shown that a large ion temperature gradient can increase the bootstrap current. When the plasma rotation is taken into account, the toroidal rotation gradient can drive an additional parallel flow for the ions and then additional bootstrap current, either positive or negative, depending on the gradient direction. Compared with the carbon radial force balance estimate for the neoclassical poloidal flow, our nonlocal simulation predicts a significantly deeper radial electric field well at the location of an internal transport barrier of an NSTX discharge

  7. Physical design of MW-class steady-state spherical tokamak, QUEST

    International Nuclear Information System (INIS)

    Hanada, K.; Sato, K.N.; Zushi, H.; Nakamura, K.; Sakamoto, M.; Idei, H.; Hasegawa, M.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Higashizono, Y.; Yoshida, N.; Takase, Y.; Ejiri, A.; Ogawa, Y.; Ono, Y.; Yoshida, Z.; Mitarai, O.; Maekawa, T.; Kishimoto, Y.; Ishiguro, M.; Yoshinaga, T.; Igami, H.; Hirooka, Y.; Komori, A.; Motojima, O.; Sudo, S.; Yamada, H.; Ando, A.; Asakura, Nobuyuki; Matsukawa, Makoto; Ishida, A.; Ohno, N.; Peng, M.

    2008-10-01

    QUEST (R=0.68 m, a=0.4 m) focuses on the steady state operation of the spherical tokamak (ST) by controlled PWI and electron Bernstain wave (EBW) current drive (CD). The QUEST project will be developed along two phases, phase I: steady state operation with plasma current, I p =20-30 kA on open divertor configuration and phase II: steady state operation with I p = 100 kA and β of 10% in short pulse on closed divertor configuration. Feasibility of the missions on QUEST was investigated and the suitable machine size of QUEST was decided based on the physical view of plasma parameters. Electron Bernstein wave (EBW) current drive are planned to establish the maintenance of plasma current in steady state. Mode conversion efficiency to EBW was calculated and the conversion of 95% will be expected. A new type antenna for QUEST has been fabricated to excite EBW effectively. The situation of heat and particle handling is challenging, and W and high temperature wall is adopted. The start-up scenario of plasma current was investigated based on the driven current by energetic electron and the most favorable magnetic configuration for start-up is proposed. (author)

  8. Investigation of compact toroid penetration for fuelling spherical tokamak plasmas on CPD

    International Nuclear Information System (INIS)

    Fukumoto, N.; Hanada, K.; Kawakami, S.

    2008-10-01

    In previous Compact Toroid (CT) injection experiments on several tokamaks, although CT fuelling had been successfully demonstrated, the CT fuelling process has been not clear yet. We have thus conducted CT injection into simple toroidal or vertical vacuum magnetic fields to investigate quantitatively dynamics of CT plasmoid in the penetration process on a spherical tokamak (ST) device. Understanding the process allows us to address appropriately one of the critical issues for practical application of CT injection on reactor-grade tokamaks. In the experiment, the CT shift amount of about 0.26 m in a vertical magnetic field has been observed by using a fast camera. In addition to toroidal magnetic field, vertical one appears to affect CT trajectory in not conventional tokamak but ST devices operated at rather low toroidal fields. We have also observed CT attacks on the target plate with an IR camera. The IR image has indicated that CT shifts 39 mm at the toroidal field of 261 G. From the calorimetric measurement, an input energy due to CT impact in vacuum without magnetic fields is also estimated to be 530 J, which agrees with the initial CT kinetic energy. (author)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  10. High-power heating experiment of spherical tokamaks by use of plasma merging

    International Nuclear Information System (INIS)

    Ueda, Yoshinobu; Ono, Yasushi

    1999-01-01

    High-power heating of spherical tokamaks (STs) has been investigated experimentally by use of plasma merging effect. When two STs were coaxially collided, thermal energy of a colliding ST was injected into a target ST during short reconnection time (Alfven time). Though the thermal energy increment increased with decreasing plasma q value, thermal energy loss during the following relaxation, tended to be smaller with increasing q. The produced high-β STs had hallower current profiles and weaker paramagnetic toroidal field than those of single STs. Those heating properties indicate the plasma merging to be a promising initial heating method of ST plasmas. (author)

  11. A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    2011-01-01

    Roč. 51, č. 8 (2011), 083050-083050 ISSN 0029-5515 R&D Projects: GA ČR GA202/08/0419; GA MŠk 7G10072 Institutional research plan: CEZ:AV0Z20430508 Keywords : spherical tokamak * electron Bernstein wave (EBW) * heating * current drive * electron cyclotron wave Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/8/083050/pdf/0029-5515_51_8_083050.pdf

  12. Ion temperature increase during MHD events on the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Ejiri, A.; Shiraiwa, S.; Takase, Y.; Yamada, T.; Nagashima, Y.; Kasahara, H.; Iijima, D.; Kobori, Y.; Nishi, T.; Taniguchi, T.; Aramasu, M.; Ohara, S.; Ushigome, M.; Yamagishi, K.

    2003-01-01

    Various types of MHD events including internal reconnection events are studied on the TST-2 spherical tokamak. In weak MHD events no positive current spike was observed, but in strong MHD events with positive current spikes, a rapid and significant impurity ion temperature increase was observed. The decrease in the poloidal magnetic energy is the most probable energy source for ion heating. The plasma current shows a stepwise change. The magnitude of this step correlates with the temperature increase and is found to be a good indicator of the strength of each event. (author)

  13. Next-Step Spherical Torus Experiment and Spherical Torus Strategy in the Fusion Energy Development Path

    International Nuclear Information System (INIS)

    Ono, M.; Peng, M.; Kessel, C.; Neumeyer, C.; Schmidt, J.; Chrzanowski, J.; Darrow, D.; Grisham, L.; Heitzenroeder, P.; Jarboe, T.; Jun, C.; Kaye, S.; Menard, J.; Raman, R.; Stevenson, T.; Viola, M.; Wilson, J.; Woolley, R.; Zatz, I.

    2003-01-01

    A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction

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

    Science.gov (United States)

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

    2015-06-01

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

  15. Plasma Turbulence Suppression and Transport Barrier Formation by Externally Driven RF Waves in Spherical Tokamaks

    International Nuclear Information System (INIS)

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

    2002-01-01

    Turbulent transport of heat and particles is the principle obstacle confronting controlled fusion today. Thus, we investigate quantitatively the suppression of turbulence and formation of transport barriers in spherical tokamaks by sheared electric fields generated by externally driven radio-frequency (RF) waves, in the frequency range o)A n o] < o)ci (e)A and o)ci are the Alfven and ion cyclotron frequencies). This investigation consists of the solution of the full-wave equation for a spherical tokamak in the presence of externally driven fast waves and the evaluation of the power dissipation by the mode-converted Alfven waves. This in turn, provides a radial flow shear responsible for the suppression of plasma turbulence. Thus, a strongly non-linear equation for the radial sheared electric field is solved, the turbulent transport suppression rate is evaluated and compared with the ion temperature gradient (ITG) instability increment. For illustration, the case of START-like device (Sykes 2000) is treated. Thus, (i) the exact D-shape cross-section is considered; (ii) additional kinetic (including Landau damping) and particle trapping effects are added to the resistive two-fluid dielectric tensor operator; (iii) a finite extension antenna located on the low-field-side of the plasma is considered; (iv) a rigorous 2.5 finite elements numerical code (Sewell 1993) is used; and (v) the turbulence and transport barrier generated as a result of wave-plasma interaction is evaluated

  16. A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

    Science.gov (United States)

    Urban, Jakub; Decker, Joan; Peysson, Yves; Preinhaelter, Josef; Shevchenko, Vladimir; Taylor, Gary; Vahala, Linda; Vahala, George

    2011-08-01

    The electron Bernstein wave (EBW) is typically the only wave in the electron cyclotron (EC) range that can be applied in spherical tokamaks for heating and current drive (H&CD). Spherical tokamaks (STs) operate generally in high-β regimes, in which the usual EC O- and X-modes are cut off. In this case, EBWs seem to be the only option that can provide features similar to the EC waves—controllable localized H&CD that can be used for core plasma heating as well as for accurate plasma stabilization. The EBW is a quasi-electrostatic wave that can be excited by mode conversion from a suitably launched O- or X-mode; its propagation further inside the plasma is strongly influenced by the plasma parameters. These rather awkward properties make its application somewhat more difficult. In this paper we perform an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX). Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions, which are the fundamental EBW parameters that can be chosen and controlled. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

  17. A modulation model for mode splitting of magnetic perturbations in the Mega Ampere Spherical Tokamak

    International Nuclear Information System (INIS)

    Hole, M J; Appel, L C

    2009-01-01

    Recent observations of magnetic fluctuation activity in the Mega Ampere Spherical Tokamak (MAST) reveal the presence of plasmas with bands of both low and high frequency magnetic fluctuations. Such plasmas exhibit a spectrum of low frequency modes with adjacent toroidal mode numbers, for which the measured frequency is near the Doppler shifted rotation frequency of the plasma. These are thought to be tearing modes. Also present are a spectrum of high frequency modes (e.g. Alfven, fishbone and/or ICE). The frequency and mode number of the tearing mode and its harmonics is identical to the frequency and mode number splitting of the high frequency MHD activity, strongly suggesting that the high frequency splitting is produced by modulation of the high and low frequency modes. We describe a strong modulation model, in which the nonlinear terms are fitted to produce the amplitude envelope profile of the tearing mode. A bispectral analysis proves that the low frequency modes are indeed in phase with the fundamental, while Fourier-SVD mode analysis confirms the mode numbers are toroidal harmonics. Employing this model, the sideband amplitude profile of the high frequency modes is predicted, and found to be in good agreement with experimental observations. Also, toroidal mode number splitting of the high frequency activity matches the mode number of the tearing mode. Weak evidence is found to indicate the Alfvenic sidebands are in phase with the Alfven eigenmode fundamental. The findings support predictions of a strong modulation model, and suggest a need to further develop nonlinear MHD theory to predict the amplitude of coupled sidebands, and so corroborate the observed nonlinear plasma response.

  18. Neutronics analysis of the conceptual design of a component test facility based on the spherical tokamak

    International Nuclear Information System (INIS)

    Zheng, S.; Voss, G.M.; Pampin, R.

    2010-01-01

    One of the crucial aspects of fusion research is the optimisation and qualification of suitable materials and components. To enable the design and construction of DEMO in the future, ITER is taken to demonstrate the scientific and technological feasibility and IFMIF will provide rigorous testing of small material samples. Meanwhile, a dedicated, small-scale components testing facility (CTF) is proposed to complement and extend the functions of ITER and IFMIF and operate in association with DEMO so as to reduce the risk of delays during this phase of fusion power development. The design of a spherical tokamak (ST)-based CTF is being developed which offers many advantages over conventional machines, including lower tritium consumption, easier maintenance, and a compact assembly. The neutronics analysis of this system is presented here. Based on a three-dimensional neutronics model generated by the interface programme MCAM from CAD models, a series of nuclear and radiation protection analyses were carried out using the MCNP code and FENDL2.1 nuclear data library to assess the current design and guide its development if needed. The nuclear analyses addresses key neutronics issues such as the neutron wall loading (NWL) profile, nuclear heat loads, and radiation damage to the coil insulation and to structural components, particularly the stainless steel vessel wall close to the NBI ports where shielding is limited. The shielding of the divertor coil and the internal Poloidal Field (PF) coil, which is introduced in the expanded divertor design, are optimised to reduce their radiation damage. The preliminary results show that the peak radiation damage to the structure of martensitic/ferritic steel is about 29 dpa at the mid-plane assuming a life of 12 years at a duty factor 33%, which is much lower than its ∼150 dpa limit. In addition, TBMs installed in 8 mid-plane ports and 6 lower ports, and 60% 6 Li enrichment in the Li 4 SiO 4 breeder, the total tritium generation is

  19. Transport and confinement in the Mega Ampere Spherical Tokamak (MAST) plasma

    International Nuclear Information System (INIS)

    Akers, R J; Ahn, J W; Antar, G Y; Appel, L C; Applegate, D; Brickley, C; Bunting, C; Carolan, P G; Challis, C D; Conway, N J; Counsell, G F; Dendy, R O; Dudson, B; Field, A R; Kirk, A; Lloyd, B; Meyer, H F; Morris, A W; Patel, A; Roach, C M; Rohzansky, V; Sykes, A; Taylor, D; Tournianski, M R; Valovic, M; Wilson, H R; Axon, K B; Buttery, R J; Ciric, D; Cunningham, G; Dowling, J; Dunstan, M R; Gee, S J; Gryaznevich, M P; Helander, P; Keeling, D L; Knight, P J; Lott, F; Loughlin, M J; Manhood, S J; Martin, R; McArdle, G J; Price, M N; Stammers, K; Storrs, J; Walsh, M J

    2003-01-01

    A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampere Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement H H factor (w.r.t. scaling law IPB98[y, 2]) around approx. 1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L-H power threshold scaling proportional to plasma surface area (rather than P LH approx. R 2 ). In addition, MAST favours an inverse aspect ratio scaling P LH approx. epsilon 0.5. Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling W ped approx. epsilon -2.13 and modifies the exponents on R, B T and Kappa. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using techniques developed at conventional aspect

  20. Transport and confinement in the Mega Ampere Spherical Tokamak (MAST) plasma

    Energy Technology Data Exchange (ETDEWEB)

    Akers, R J; Ahn, J W; Appel, L C; Brickley, C; Bunting, C; Carolan, P G; Challis, C D; Conway, N J; Counsell, G F; Dendy, R O; Dudson, B; Field, A R; Kirk, A; Lloyd, B; Meyer, H F; Morris, A W; Patel, A; Roach, C M; Sykes, A; Taylor, D; Tournianski, M R; Valovic, M; Wilson, H R; Axon, K B; Buttery, R J; Ciric, D; Cunningham, G; Dowling J; Dunstan, M R; Gee, S J; Gryaznevich, M P; Helander, P; Keeling, D L; Knight, P J; Lott, F; Loughlin, M J; Manhood, S J; Martin, R; McArdle, G J; Price, M N; Stammers, K; Storrs, J [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Antar, G Y [Fusion Energy Research Program, University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Applegate, D [Imperial College of Science, Technology and Medicine, University of London, London SW7 2BZ (United Kingdom); Rohzansky, V [St. Petersburg State Politechnical University, Polytechnicheskaya 29, 195251 St. Petersburg (Russian Federation); Walsh, M J [Walsh Scientific Ltd., Abingdon, Oxon OX14 3EB (United Kingdom)

    2003-12-01

    A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampere Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement H{sub H} factor (w.r.t. scaling law IPB98[y, 2]) around approx. 1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L-H power threshold scaling proportional to plasma surface area (rather than P{sub LH} approx. R{sup 2}). In addition, MAST favours an inverse aspect ratio scaling P{sub LH} approx. epsilon 0.5. Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling W{sub ped} approx. epsilon -2.13 and modifies the exponents on R, B{sub T} and Kappa. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using

  1. Preliminary experiment of non-induced plasma current startup on SUNIST spherical tokamak

    International Nuclear Information System (INIS)

    He Yexi; Zhang Liang; Xie Lifeng; Tang Yi; Yang Xuanzong; Fu Hongjun

    2005-01-01

    Non-inductive plasma current startup is an important motivation on the SUNIST spherical tokamak. In this experiment, a 100 kW, 2.45 GHz magnetron microwave system has been applied to the plasma current startup. Besides the toroidal field, a vertical field was applied to generate a preliminary toroidal plasma current without action of the central solenoid. As the evidence of the plasma current startup by the vertical field drift effect, the direction of the plasma current is changed with the changing direction of the vertical field during ECR startup discharge. We have also observed the plasma current maximum by scanning the vertical field in both directions. Additionally, we have used electrode discharge to assist the ECR current startup. (author)

  2. Plasma heating and fuelling in the Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Barsukov, A.G.; Belyakov, V.A.

    2005-01-01

    The results of the last two years of plasma investigations at Globus-M are presented. Described are improvements helping to achieve high performance OH plasmas, which are used as the target for auxiliary heating and fuelling experiments. Increased energy content, high beta poloidal and good confinement are reported. Experiments on NBI plasma heating with a wide range of plasma parameters were performed. Some results are presented and analyzed. Experiments on RF plasma heating in the frequency range of fundamental ion cyclotron harmonics are described. In some experiments which were performed for the first time in spherical tokamaks, promising results were achieved. Noticeable ion heating was recorded at low launched power and a high concentration of hydrogen minority in deuterium plasmas. Simulations of RF wave absorption are briefly discussed. Described also are modification of the plasma gun and test-stand experiments. Fuelling experiments performed at Globus-M are discussed. (author)

  3. RF start-up and sustainment experiments on the TST-2-K spherical tokamak

    International Nuclear Information System (INIS)

    Ejiri, A.; Takase, Y.; Kasahara, H.; Yamada, T.; Hanada, K.; Sato, K. N.; Zushi, H.; Nakamura, K.; Sakamoto, M.; Idei, H.; Hasegawa, M.; Iyomasa, A.; Imamura, N.; Esaki, K.; Kitaguchi, M.; Sasaki, K.; Hoshika, H.; Mitarai, O.; Nishino, N.

    2006-01-01

    Plasma start-up and sustainment without an inductive field have been studied in the TST-2-K spherical tokamak using high power RF sources (8.2 GHz/up to 170 kW). Steady state discharges with a plasma current of 4 kA were achieved. The line integrated density was about 3 x 10 17 m -2 and the electron temperature was 160 eV. A truncated equilibrium was introduced to reproduce magnetic measurements. It was found that a positive Pfirsch-Schlueter current in the open field line region at the outboard boundary makes a significant contribution to the current. Insensitivity of the current to variations in the vertical field and RF power variation was also found

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

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Sabbagh, S.A.; Fredrickson, E.D.; Jardin, S.C.; Maingi, R.; Manickam, J.; Mueller, D.; Ono, M.; Paoletti, F.; Peng, Y.-K.M.; Soukhanovskii, V.; Stutman, D.; Synakowski, E.J.

    2003-01-01

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

  5. Transmutation of minor actinides in a spherical torus tokamak fusion reactor, FDTR

    International Nuclear Information System (INIS)

    Feng, K.M.; Zhang, G.S.; Deng, M.G.

    2003-01-01

    In this paper, a concept for the transmutation of minor actinide (MA) nuclear wastes based on a spherical torus (ST) tokamak reactor, FDTR, is put forward. A set of plasma parameters suitable for the transmutation blanket was chosen. The 2-D neutron transport code TWODANT, the 3-D Monte Carlo code MCNP/4B, the 1-D neutron transport and burn-up calculation code BISON3.0 and their associated data libraries were used to calculate the transmutation rate, the energy multiplication factor and the tritium breeding ratio of the transmutation blanket. The calculation results for the system parameters and the actinide series isotopes for different operation times are presented. The engineering feasibility of the center-post (CP) of FDTR has been investigated and the results are also given. A preliminary neutronics calculation based on an ST transmutation blanket shows that the proposed system has a high transmutation capability for MA wastes. (author)

  6. Analysis and design of the Alfven wave antenna system for the SUNIST spherical tokamak

    International Nuclear Information System (INIS)

    Tan Yi; Gao Zhe; He Yexi

    2009-01-01

    Analysis and design of the Alfven wave antenna system for the SUNIST spherical tokamak are presented. Two candidate antenna concepts, folded and unfolded, are analyzed and compared with each other. In the frequency range of Alfven resonance the impedance spectrums of both two concept antennas for major modes are numerically calculated in a 1-D MHD framework. The folded concept is chosen for engineering design. The antenna system is designed to be simple and requires least modification to the vacuum vessel. The definition of the antenna shape is guided by the analyses with constraints of existing hardware layouts. Each antenna unit consists of two stainless steel straps with a thickness of 1 mm. A number of boron nitride tiles are assembled together as the side limiters for plasma shielding. Estimation shows that the structure is robust enough to withstand the electromagnetic force and the heat load for typical discharge duty cycles.

  7. Physics of energetic particle-driven instabilities in the START spherical tokamak

    International Nuclear Information System (INIS)

    McClements, K.G.; Gryaznevich, M.P.; Akers, R.J.; Appel, L.C.; Counsell, G.F.; Roach, C.M.; Sharapov, S.E.; Majeski, R.

    1999-01-01

    The recent use of neutral beam injection (NBI) in the UKAEA small tight aspect ratio tokamak (START) has provided the first opportunity to study experimentally the physics of energetic ions in spherical tokamak (ST) plasmas. In such devices the ratio of major radius to minor radius R 0 /a is of order unity. Several distinct classes of NBI-driven instability have been observed at frequencies up to 1 MHz during START discharges. These observations are described, and possible interpretations are given. Equilibrium data, corresponding to times of beam-driven wave activity, are used to compute continuous shear Alfven spectra: toroidicity and high plasma beta give rise to wide spectral gaps, extending up to frequencies of several times the Alfven gap frequency. In each of these gaps Alfvenic instabilities could, in principle, be driven by energetic ions. Chirping modes observed at high beta in this frequency range have bandwidths comparable to or greater than the gap widths. Instability drive in START is provided by beam ion pressure gradients (as in conventional tokamaks), and also by positive gradients in beam ion velocity distributions, which arise from velocity-dependent charge exchange losses. It is shown that fishbone-like bursts observed at a few tens of kHz can be attributed to internal kink mode excitation by passing beam ions, while narrow-band emission at several hundred kHz may be due to excitation of fast Alfven (magnetosonic) eigenmodes. In the light of our understanding of energetic particle-driven instabilities in START, the possible existence of such instabilities in larger STs is discussed. (author)

  8. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe

    International Nuclear Information System (INIS)

    Walkden, N. R.; Adamek, J.; Komm, M.; Allan, S.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Dudson, B. D.

    2015-01-01

    The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ∼1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the E R measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak

  9. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe

    Science.gov (United States)

    Walkden, N. R.; Adamek, J.; Allan, S.; Dudson, B. D.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Komm, M.

    2015-02-01

    The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ˜1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the ER measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak.

  10. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe

    Energy Technology Data Exchange (ETDEWEB)

    Walkden, N. R., E-mail: nrw504@york.ac.uk [CCFE, Culham Science Centre, Abingdon,Oxon OX14 3DB (United Kingdom); Department of Physics, York Plasma Institute, University of York, Heslington, York YO10 5DD (United Kingdom); Adamek, J.; Komm, M. [Institute of Plasma Physics of AS CR, v. v. i., Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Allan, S.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A. [CCFE, Culham Science Centre, Abingdon,Oxon OX14 3DB (United Kingdom); Dudson, B. D. [Department of Physics, York Plasma Institute, University of York, Heslington, York YO10 5DD (United Kingdom)

    2015-02-15

    The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ∼1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the E{sub R} measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak.

  11. Electron cyclotron heating/current-drive system using high power tubes for QUEST spherical tokamak

    Science.gov (United States)

    Onchi, Takumi; Idei, H.; Hasegawa, M.; Nagata, T.; Kuroda, K.; Hanada, K.; Kariya, T.; Kubo, S.; Tsujimura, T. I.; Kobayashi, S.; Quest Team

    2017-10-01

    Electron cyclotron heating (ECH) is the primary method to ramp up plasma current non-inductively in QUEST spherical tokamak. A 28 GHz gyrotron is employed for short pulses, where the radio frequency (RF) power is about 300 kW. Current ramp-up efficiency of 0.5 A/W has been obtained with focused beam of the second harmonic X-mode. A quasi-optical polarizer unit has been newly installed to avoid arcing events. For steady-state tokamak operation, 8.56 GHz klystron with power of 200 kW is used as the CW-RF source. The high voltage power supply (54 kV/13 A) for the klystron has been built recently, and initial bench test of the CW-ECH system is starting. The array of insulated-gate bipolar transistor works to quickly cut off the input power for protecting the klystron. This work is supported by JSPS KAKENHI (15H04231), NIFS Collaboration Research program (NIFS13KUTR085, NIFS17KUTR128), and through MEXT funding for young scientists associated with active promotion of national university reforms.

  12. Developement of Spherical Polyurethane Beads

    Institute of Scientific and Technical Information of China (English)

    K. Maeda; H. Ohmori; H. Gyotoku

    2005-01-01

    @@ 1Results and Discussion We established a new method to produce the spherical polyurethane beads which have narrower distribution of particle size. This narrower distribution was achieved by the polyurethane prepolymer which contains ketimine as a blocked chain-extending agent. Firstly, the prepolymer is dispersed into the aqueous solution containing surfactant. Secondaly, water comes into the inside of prepolymer as oil phase. Thirdly, ketimine is hydrolyzed to amine, and amine reacts with prepolymer immediately to be polyurethane.Our spherical polyurethane beads are very suitable for automotive interior parts especially for instrument panel cover sheet producing under the slush molding method, because of good process ability, excellent durability to the sunlight and mechanical properties at low temperature. See Fig. 1 ,Fig. 2 and Fig. 3 (Page 820).

  13. Design innovations of the next-step spherical torus experiment and spherical torus development path

    International Nuclear Information System (INIS)

    Ono, M.; Kessel, C.; Peng, M.

    2003-01-01

    The spherical torus (ST) fusion energy development path is complementary to the tokamak burning plasma experiment such as ITER as it focuses toward the compact Component Test Facility (CTF) and higher toroidal beta regimes to improve the design of DEMO and a Power Plant. To support the ST development path, one option of a Next Step Spherical Torus (NSST) device is examined. NSST is a 'performance extension' (PE) stage ST with a plasma current of 5 - 10 MA, R = 1.5, B T ≤ 2.7 T with flexible physics capability to 1) Provide a sufficient physics basis for the design of the CTF, 2) Explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, 3) Contribute to the general plasma/fusion science of high β toroidal plasmas. The NSST facility is designed to utilize the TFTR site to minimize the cost and time required for the construction. (author)

  14. Proposed tokamak poloidal field system development program

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, J.D.; Vogel, H.F.; Warren, R.W.; Weldon, D.M.

    1977-05-01

    A program is proposed to develop poloidal field components for TNS and EPR size tokamak devices and to test these components in realistic circuits. Emphasis is placed upon the development of the most difficult component, the superconducting ohmic-heating coil. Switches must also be developed for testing the coils, and this switching technology is to be extended to meet the requirements for the large scale tokamaks. Test facilities are discussed; power supplies, including a homopolar to drive the coils, are considered; and poloidal field systems studies are proposed.

  15. Collisional Damping of Electron Bernstein Waves and its Mitigation by Evaporated Lithium Conditioning in Spherical-Tokamak Plasmas

    International Nuclear Information System (INIS)

    Diem, S. J.; Caughman, J. B.; Taylor, G.; Efthimion, P. C.; Kugel, H.; LeBlanc, B. P.; Phillips, C. K.; Preinhaelter, J.; Urban, J.; Sabbagh, S. A.

    2009-01-01

    The first experimental verification of electron Bernstein wave (EBW) collisional damping, and its mitigation by evaporated Li conditioning, in an overdense spherical-tokamak plasma has been observed in the National Spherical Torus Experiment (NSTX). Initial measurements of EBW emission, coupled from NSTX plasmas via double-mode conversion to O-mode waves, exhibited <10% transmission efficiencies. Simulations show 80% of the EBW energy is dissipated by collisions in the edge plasma. Li conditioning reduced the edge collision frequency by a factor of 3 and increased the fundamental EBW transmission to 60%.

  16. Developments in tokamak transport modeling

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  17. Linear stability and nonlinear dynamics of the fishbone mode in spherical tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Feng; Liu, J. Y. [School of Physics and Optoelectronic Engineering, Dalian University of Technology, Dalian 116024 (China); Fu, G. Y.; Breslau, J. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2013-10-15

    Extensive linear and nonlinear simulations have been carried out to investigate the energetic particle-driven fishbone instability in spherical tokamak plasmas with weakly reversed q profile and the q{sub min} slightly above unity. The global kinetic-MHD hybrid code M3D-K is used. Numerical results show that a fishbone instability is excited by energetic beam ions preferentially at higher q{sub min} values, consistent with the observed appearance of the fishbone before the “long-lived mode” in MAST and NSTX experiments. In contrast, at lower q{sub min} values, the fishbone tends to be stable. In this case, the beam ion effects are strongly stabilizing for the non-resonant kink mode. Nonlinear simulations show that the fishbone saturates with strong downward frequency chirping as well as radial flattening of the beam ion distribution. An (m, n) = (2, 1) magnetic island is found to be driven nonlinearly by the fishbone instability, which could provide a trigger for the (2, 1) neoclassical tearing mode sometimes observed after the fishbone instability in NSTX.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  19. A high resolution Mirnov array for the Mega Ampere Spherical Tokamak

    International Nuclear Information System (INIS)

    Hole, M. J.; Appel, L. C.; Martin, R.

    2009-01-01

    Over the past two decades, the increase in neutral-beam heating and α particle production in magnetically confined fusion plasmas has led to an increase in energetic particle driven mode activity, much of which has an electromagnetic signature which can be detected by the use of external Mirnov coils. Typically, the frequency and spatial wave number band of such oscillations increase with increasing injection energy, offering new challenges for diagnostic design. In particular, as the frequency approaches the megahertz range, care must be taken to model the stray capacitance of the coil, which limits the resonant frequency of the probe; model transmission line effects in the system, which if unchecked can produce system resonances; and minimize coil conductive shielding, so as to minimize skin currents which limit the frequency response of the coil. As well as optimizing the frequency response, the coils should also be positioned to confidently identify oscillations over a wide wave number band. This work, which draws on new techniques in stray capacitance modeling and coil positioning, is a case study of the outboard Mirnov array for high-frequency acquisition in the Mega Ampere Spherical Tokamak, and is intended as a roadmap for the design of high frequency, weak field strength magnetic diagnostics.

  20. Integrated predictive modeling simulations of the Mega-Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.; Akers, Robert; Byrom, Calum; Sykes, Alan

    2002-01-01

    Integrated predictive modeling simulations are carried out using the BALDUR transport code [Singer et al., Comput. Phys. Commun. 49, 275 (1982)] for high confinement mode (H-mode) and low confinement mode (L-mode) discharges in the Mega-Amp Spherical Tokamak (MAST) [Sykes et al., Phys. Plasmas 8, 2101 (2001)]. Simulation results, obtained using either the Multi-Mode transport model (MMM95) or, alternatively, the mixed-Bohm/gyro-Bohm transport model, are compared with experimental data. In addition to the anomalous transport, neoclassical transport is included in the simulations and the ion thermal diffusivity in the inner third of the plasma is found to be predominantly neoclassical. The sawtooth oscillations in the simulations radially spread the neutral beam injection heating profiles across a broad sawtooth mixing region. The broad sawtooth oscillations also flatten the central temperature and electron density profiles. Simulation results for the electron temperature and density profiles are compared with experimental data to test the applicability of these models and the BALDUR integrated modeling code in the limit of low aspect ratio toroidal plasmas

  1. 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

  2. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

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

  3. Plasma jet source parameter optimisation and experiments on injection into Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Petrov, Yu.V.; Sakharov, N.V.; Semenov, A.A.; Voronin, A.V.

    2005-01-01

    Results of theoretical and experimental research on the plasma sources and injection of plasma and gas jet produced by the modified source into tokamak Globus-M are presented. An experimental test stand was developed for investigation of intense plasma jet generation. Optimisation of pulsed coaxial accelerator parameters by means of analytical calculations is performed with the aim of achieving the highest flow velocity at limited coaxial electrode length and discharge current. The optimal parameters of power supply to generate a plasma jet with minimal impurity contamination and maximum flow velocity were determined. A comparison of experimental and calculation results is made. Plasma jet parameters are measured, such as: impurity species content, pressure distribution across the jet, flow velocity, plasma density, etc. Experiments on the interaction of a higher kinetic energy plasma jet with the magnetic field and plasma of the Globus-M tokamak were performed. Experimental results on plasma and gas jet injection into different Globus-M discharge phases are presented and discussed. Results are presented on the investigation of plasma jet injection as the source for discharge breakdown, plasma current startup and initial density rise. (author)

  4. High-beta characteristics of first and second-stable spherical tokamaks in reconnection heating experiments of TS-3

    International Nuclear Information System (INIS)

    Ono, Y.

    2002-01-01

    Novel formations of ultra-high-beta Spherical Tokamak (ST) have been developed in the TS-3 device using high power heating of merging/ reconnection. In Type-A merging, two STs were merged together to build up the plasma beta. In Type-B merging, an oblate FRC was initially formed by merging of two spheromaks with opposing toroidal field B t and was transformed into an ultra-high-beta ST by applying external B t . Ballooning stability analyses confirmed formations of the first-stable STs by Type- A merging and the second-stable STs by Type-B merging and also the unstable STs by both mergings, revealing the ballooning stability window consistent with measured high-n instabilities. We made (1) those model analyses of the produced STs for the first time using the BALLOO stability code, revealing that hollowness/ broadness of current/pressure profiles widen significantly the window to the second-stable regime. This paper also addresses (2) normalized betas of the second-stable STs as large as 6-17 for comparison with the Troyon scaling and (3) a promising scaling of the reconnection heating energy. (author)

  5. Particle acceleration during merging-compression plasma start-up in the Mega Amp Spherical Tokamak

    Science.gov (United States)

    McClements, K. G.; Allen, J. O.; Chapman, S. C.; Dendy, R. O.; Irvine, S. W. A.; Marshall, O.; Robb, D.; Turnyanskiy, M.; Vann, R. G. L.

    2018-02-01

    Magnetic reconnection occurred during merging-compression plasma start-up in the Mega Amp Spherical Tokamak (MAST), resulting in the prompt acceleration of substantial numbers of ions and electrons to highly suprathermal energies. Accelerated field-aligned ions (deuterons and protons) were detected using a neutral particle analyser at energies up to about 20 keV during merging in early MAST pulses, while nonthermal electrons have been detected indirectly in more recent pulses through microwave bursts. However no increase in soft x-ray emission was observed until later in the merging phase, by which time strong electron heating had been detected through Thomson scattering measurements. A test-particle code CUEBIT is used to model ion acceleration in the presence of an inductive toroidal electric field with a prescribed spatial profile and temporal evolution based on Hall-MHD simulations of the merging process. The simulations yield particle distributions with properties similar to those observed experimentally, including strong field alignment of the fast ions and the acceleration of protons to higher energies than deuterons. Particle-in-cell modelling of a plasma containing a dilute field-aligned suprathermal electron component suggests that at least some of the microwave bursts can be attributed to the anomalous Doppler instability driven by anisotropic fast electrons, which do not produce measurable enhancements in soft x-ray emission either because they are insufficiently energetic or because the nonthermal bremsstrahlung emissivity during this phase of the pulse is below the detection threshold. There is no evidence of runaway electron acceleration during merging, possibly due to the presence of three-dimensional field perturbations.

  6. The poloidal distribution of turbulent fluctuations in the Mega-Ampere Spherical Tokamak

    International Nuclear Information System (INIS)

    Antar, G.Y.; Counsell, G.; Ahn, J.-W.; Yang, Y.; Price, M.; Tabasso, A.; Kirk, A.

    2005-01-01

    Recently, it was shown that intermittency observed in magnetic fusion devices is caused by large-scales events with high radial velocity reaching about 1/10th of the sound speed (called avaloids or blobs) [G. Antar et al., Phys. Rev. Lett. 87 065001 (2001)]. In the present paper, the poloidal distribution of turbulence is investigated on the Mega-Ampere Spherical Tokamak [A. Sykes et al., Phys. Plasmas 8 2101 (2001)]. To achieve our goal, target probes that span the divertor strike points are used and one reciprocating probe at the midplane. Moreover, a fast imaging camera that can reach 10 μs exposure time looks tangentially at the plasma allowing us to view a poloidal cut of the plasma. The two diagnostics allow us to have a rather accurate description of the particle transport in the poloidal plane for L-mode discharges. Turbulence properties at the low-field midplane scrape-off layer are discussed and compared to other poloidal positions. On the low-field target divertor plates, avaloids bursty signature is not detected but still intermittency is observed far from the strike point. This is a consequence of the field line expansion which transforms a structure localized in the poloidal plane into a structure which expands over several tens of centimeters at the divertor target plates. Around the X point and in the high-field side, however, different phenomena enter into play suppressing the onset of convective transport generation. No signs of intermittency are observed in these regions. Accordingly, like 'normal' turbulence, the onset of convective transport is affected by the local magnetic curvature and shear

  7. Effects of an Anomalous Resistivity on the Power Deposition by Alfven Waves in Pre-Heated Spherical Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Bruma, C.; Cuperman, S.; Komoshvili, K. [Tel Aviv Univ., Ramat Aviv (Israel)

    2005-08-01

    As it is the case with tokamaks in general, and moreover, due to their specific geometry (limited space for inboard solenoid magnets), low aspect ratio (spherical) tokamaks (STs) require additional auxiliary non-ohmic current startup and maintenance, generation of internal transport barriers (associated with underlying sheared poloidal flows and quasi-stationary radial electric fields), plasma heating, etc. One of the options to generate these necessary effects in STs is by the aid of rf waves launched from a suitable external antenna; in this option the effects just mentioned are a consequence of ponderomotive forces resulting from the interaction of the rf waves with the plasma. Since experimental data on STs (viz., the START-device) reveal the presence of an anomalous plasma resistivity (about four times Spitzer's one), we carried out a systematic parametric investigation of the effects of an increased plasma resistivity on the magnitude and spatial localization of the resulting power deposition.

  8. Tokamak first-wall coating program development

    International Nuclear Information System (INIS)

    Davis, M.J.; Langley, R.A.; Prevender, T.S.

    1977-08-01

    The development of a research program to study coatings for control of impurities originating from the first wall of a Tokamak reactor is extensively discussed. The first wall environment and sputtering, temperature, surface chemical, and bulk radiation damage effects are reviewed. Candidate materials and application techniques are discussed. The philosophy and flow chart of a recommended coating development plan are presented and discussed. Projected impacts of the proposed plan include benefits to other aspects of confinement experiments. A list of 45 references is appended

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

    International Nuclear Information System (INIS)

    Sinman, S.; Sinman, A.

    2003-01-01

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

  10. Non-inductive plasma initiation and plasma current ramp-up on the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Takase, Y.; Ejiri, A.; Oosako, T.; Shinya, T.; Ambo, T.; Furui, H.; Kato, K.; Nakanishi, A.; Sakamoto, T.; Kakuda, H.; Wakatsuki, T.; Hashimoto, T.; Hiratsuka, J.; Kasahara, H.; Kumazawa, R.; Mutoh, T.; Saito, K.; Seki, T.; Moeller, C.P.; Nagashima, Y.

    2013-01-01

    Plasma current (I p ) start-up in a spherical tokamak (ST) by waves in the lower-hybrid (LH) frequency range was investigated on TST-2. A low current (∼1 kA) ST configuration can be formed by waves over a broad frequency range (21 MHz–8.2 GHz in TST-2), but further I p ramp-up (to ∼10 kA) is most efficient with waves in the LH frequency range. I p ramp-up to 15 kA was achieved with 60 kW of net RF power P RF in the fast wave (FW) polarization at 200 MHz excited by the inductively coupled combline antenna. X-ray measurements showed that the photon flux and temperature are higher in the direction opposite to I p , consistent with acceleration of electrons by a uni-directional RF wave. There is evidence that the LH wave is excited nonlinearly by the FW, based on the frequency spectra measured by magnetic probes. Similar efficiencies of I p ramp-up were obtained with the inductive combline antenna and the dielectric-loaded waveguide array (‘grill’) antenna, and tendencies for the current drive efficiency to increase with plasma current and toroidal field were observed. During operation of the grill antenna, wavevector components were measured by an array of magnetic probes. Results were qualitatively consistent with expectations based on dispersion relations for the FW and the LH wave. A capacitively coupled combline antenna has been developed to improve coupling to the plasma and the wavenumber spectrum of the excited LH wave, and will be tested in 2013. (paper)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  12. Development of atomic beam probe for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M., E-mail: bertam@sze.hu [Széchenyi István University, EURATOM Association, Győr (Hungary); Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, EURATOM Association, Budapest (Hungary); Havlícek, J.; Háček, P. [Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics (Czech Republic)

    2013-11-15

    Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies.

  13. Development of atomic beam probe for tokamaks

    International Nuclear Information System (INIS)

    Berta, M.; Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S.; Havlícek, J.; Háček, P.

    2013-01-01

    Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies

  14. Experimental study of parametric dependence of electron-scale turbulence in a spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Y.; Guttenfelder, W.; Kaye, S. M.; Mazzucato, E.; Bell, R. E.; Diallo, A.; LeBlanc, B. P. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Domier, C. W.; Lee, K. C. [University of California at Davis, Davis, California 95616 (United States); Smith, D. R. [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Yuh, H. [Nova Photonics, Inc., Princeton, New Jersey 08540 (United States)

    2012-05-15

    Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak Experiment (NSTX), electron-scale density fluctuations are studied with a novel tangential microwave scattering system with high radial resolution of {+-}2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an edge localized mode (ELM) event, where we found the first clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, coupled with linear gyro-kinetic calculations, leads to the identification of the observed instability as toroidal electron temperature gradient (ETG) modes. It is observed that longer wavelength ETG modes, k{sub Up-Tack }{rho}{sub s} Less-Than-Or-Equivalent-To 10 ({rho}{sub s} is the ion gyroradius at electron temperature and k{sub Up-Tack} is the wavenumber perpendicular to local equilibrium magnetic field), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal diffusivity. Comparisons with nonlinear ETG gyrokinetic simulations show ETG turbulence may be able to explain the experimental electron heat flux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal field, so that electron gyroradius ({rho}{sub e}), electron beta ({beta}{sub e}), and safety factor (q{sub 95}) are kept approximately constant. More than a factor of two change in electron collisionality, {nu}{sub e}{sup *}, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as {nu}{sub e}{sup *} is

  15. 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

  16. 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...

  17. 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.

  18. Investigating fusion plasma instabilities in the Mega Amp Spherical Tokamak using mega electron volt proton emissions (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Perez, R. V., E-mail: rvale006@fiu.edu; Boeglin, W. U.; Angulo, A.; Avila, P.; Leon, O.; Lopez, C. [Department of Physics, Florida International University, 11200 SW 8 ST, CP204, Miami, Florida 33199 (United States); Darrow, D. S. [Princeton Plasma Physics Laboratory, James Forrestal Campus, P.O. Box 451, Princeton, New Jersey 08543 (United States); Cecconello, M.; Klimek, I. [Department of Physics and Astronomy, Uppsala University, Uppsala SE-751 20 (Sweden); Allan, S. Y.; Akers, R. J.; Keeling, D. L.; McClements, K. G.; Scannell, R.; Conway, N. J. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Turnyanskiy, M. [ITER Physics Department, EFDA CSU Garching, Boltzmannstrasse 2, D-85748, Garching (Germany); Jones, O. M. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Department of Physics, Durham University, Durham DH1 3LE (United Kingdom); Michael, C. A. [Australian National University, Canberra ACT 0200 (Australia)

    2014-11-15

    The proton detector (PD) measures 3 MeV proton yield distributions from deuterium-deuterium fusion reactions within the Mega Amp Spherical Tokamak (MAST). The PD’s compact four-channel system of collimated and individually oriented silicon detectors probes different regions of the plasma, detecting protons (with gyro radii large enough to be unconfined) leaving the plasma on curved trajectories during neutral beam injection. From first PD data obtained during plasma operation in 2013, proton production rates (up to several hundred kHz and 1 ms time resolution) during sawtooth events were compared to the corresponding MAST neutron camera data. Fitted proton emission profiles in the poloidal plane demonstrate the capabilities of this new system.

  19. Coherence imaging of scrape-off-layer and divertor impurity flows in the Mega Amp Spherical Tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Silburn, S. A., E-mail: s.a.silburn@durham.ac.uk; Sharples, R. M. [Centre for Advanced Instrumentation, Department of Physics, Durham University, Durham DH1 3LE (United Kingdom); Harrison, J. R.; Meyer, H.; Michael, C. A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Howard, J. [Plasma Research Laboratory, Australian National University, Canberra, ACT 0200 (Australia); Gibson, K. J. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

    2014-11-15

    A new coherence imaging Doppler spectroscopy diagnostic has been deployed on the UK’s Mega Amp Spherical Tokamak for scrape-off-layer and divertor impurity flow measurements. The system has successfully obtained 2D images of C III, C II, and He II line-of-sight flows, in both the lower divertor and main scrape-off-layer. Flow imaging has been obtained at frame rates up to 1 kHz, with flow resolution of around 1 km/s and spatial resolution better than 1 cm, over a 40° field of view. C III data have been tomographically inverted to obtain poloidal profiles of the parallel impurity flow in the divertor under various conditions. In this paper we present the details of the instrument design, operation, calibration, and data analysis as well as a selection of flow imaging results which demonstrate the diagnostic's capabilities.

  20. Ignition curves for deuterium/helium-3 fuel in spherical tokamak ...

    Indian Academy of Sciences (India)

    have been obtained in ne, T plane and, to determine the thermal instability of ... economic, environmental and safety characteristics is more attractive than an advanced ... spherical torus experiments, a magnetohydrodynamics stable high beta ...

  1. Tokamak Physics Experiment (TPX) power supply design and development

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe

    Czech Academy of Sciences Publication Activity Database

    Walkden, N.R.; Adámek, Jiří; Allan, S.; Dudson, B.D.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Komm, Michael

    2015-01-01

    Roč. 86, č. 2 (2015), č. článku 023510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * ball pen probe Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.336, year: 2015 http://dx.doi.org/10.1063/1.4908572

  3. Research and development of JT-60 tokamak

    International Nuclear Information System (INIS)

    Saito, Ryusei; Sato, Hiroshi; Murata, Toshifumi; Ito, Yoshiyasu.

    1978-01-01

    The development of nuclear fusion apparatuses for the purpose of utilizing energy due to nuclear fusion reaction has been forwarded in various countries, and in Japan, the critical plasma testing apparatus JT-60 is about to be constructed, centering around Japan Atomic Energy Research Institute. This is one of four large apparatuses to be constructed in the world, and it is expected to be completed in 1982. JT-60 is a nuclear fusion apparatus of tokamak type aiming at generating critical plasma. The features of JT-60 are the formation of the plasma with small aspect ratio, the equipment of a magnetic limiter, the arrangement of the first wall of molybdenum and high temperature baking as the measures to impurities. The large toroidal magnetic field coil of JT-60 is composed of 18 unit coils. The analyses of magnetic field, thermal behavior and structural strength, the selection of materials, and the development of manufacturing techniques regarding the toroidal coil are described. The vacuum container of JT-60 is composed of the main body of torus type comprising thickwalled rings and bellows, the first wall comprising liners, fixed limiter and magnetic limiter, and observation ports. It is large torus-form container with non-circular cross section, and baking at 500 deg. C is required as the measure to ultrahigh vacuum. Complex forces including electromagnetic force act on it. (Kako, I.)

  4. Optimization of spherical ionization chambers for neutron diagnostics in Tokamak plants

    International Nuclear Information System (INIS)

    Hoenen, F.

    1983-05-01

    For the investigation of neutron emission from fusion plasmas Pulse-Ion-Chamber are favored because of their high temporal resolution, the availability of results immedately after the discharge and their insensitivity to hard X-rays. However to measure ion temperatures below 2 keV with the aid of neutron spectroscopy the detectors have to be improved. Difficulties arise from the fact that in Pulse-Ion-Chambers the pulse height is a function of the position in the chamber where the ion pairs are produced (Induction effect). It will be shown that the induction effect is smaller in spherical ionisation chambers than in cylindrical ones. This means an increase in energy resolution so that neutrons from the D(D,n) 3 He reaction can be analysed with an energy resolution of better than 3% in spherical chambers. (orig./HP) [de

  5. On the design and role of passive stabilisation within the ST40 spherical tokamak

    Science.gov (United States)

    Buxton, P. F.; Asunta, O.; Gryaznevich, M. P.; Lockley, D.; McNamara, S.; Medvedev, S.; Ruiz de Villa Valdés, E.; Whitfield, G.; Wood, J. M.

    2018-06-01

    The position of passive stabilisation has been optimised for the low aspect ratio tokamak ST40. We find that passive stabilisation is most effective when conductors are placed near the plasma’s x-point, and the combined effect of having both inboard and outboard passive stabilisation significantly reduces the vertical instability growth rate. The growth rate can be further decreased by cooling the passive conductors down to 80 K. Two concepts for passive stabilisation are considered, passive plates and passive coils, and the relative advantages and disadvantages of each are discussed. Both concepts involve connecting the upper and lower conductors in an ‘anti-symmetric’ manner, which prevents large currents from being induced.

  6. Development of a spherical neutron rem monitor

    International Nuclear Information System (INIS)

    Panchal, C.G.; Madhavi, V.; Bansode, P.Y.; Jakati, R.K.; Ghodgaonkar, M.D.; Desai, S.S.; Shaikh, A.M.; Sathian, V.

    2007-01-01

    A new neutron rem monitor based on spherical LINUS with the state of art electronic circuits has been designed in Electronics Division. This prototype instrument encompasses a spherical double polythene moderator to improve an isotropic response and a lead layer to extend its energy response compared to the conventional neutron rem monitors. A systematic testing and calibration of the energy and directional response of the prototype monitor have been carried out. Although the monitor is expected to perform satisfactorily upto an energy ∼ 55 MeV, at present its response has been tested upto 5 MeV. (author)

  7. The influence of gas fuelling location on H-mode access in the MAST spherical tokamak

    International Nuclear Information System (INIS)

    Field, A R; Carolan, P G; Conway, N J; Counsell, G F; Cunningham, G; Helander, P; Meyer, H; Taylor, D; Tournianski, M R; Walsh, M J

    2004-01-01

    The observation that high-field side (HFS) gas puff refuelling facilitates access to the improved confinement (H-mode) regime on the COMPASS-D and MAST tokamaks prompted a theoretical investigation of the role of the neutral gas dynamics in controlling the edge plasma rotation and radial E-field, E r . Within the framework of neo-classical theory, higher edge plasma flow, and hence E r , are predicted when fuelling from the HFS-rather than from the more usual low-field side (LFS)-provided neutral viscosity dominates the transport of toroidal angular momentum. Here, these predictions are compared with experiments on MAST, where the influence of the gas-puff location on the edge E r profile is measured spectroscopically. An increase in E r is indeed observed with HFS refuelling in the region where the edge transport barrier forms, provided the neutral density at the LFS is sufficiently low so as not to damp the toroidal flow

  8. Plasma shape reconstruction of merging spherical tokamak based on modified CCS method

    Science.gov (United States)

    Ushiki, Tomohiko; Inomoto, Michiaki; Itagaki, Masafumi; McNamara, Steven

    2017-10-01

    The merging start-up method is the one of the CS-free start-up schemes that has the advantage of high plasma temperature and density because it involves reconnection heating and compression processes. In order to achieve optimal merging operations, the initial two STs should have identical plasma currents and shapes, and then move symmetrically toward the center of the device with appropriate velocity. Furthermore, from the viewpoint of the compression effect, controlling the plasma major radius is also important. To realize the active feedback control of the plasma currents, the positions, and the shapes of the two initial STs and to optimize the plasma parameters described above, accurate estimation of the plasma boundary shape is highly important. In the present work, the Modified-CCS method is demonstrated to reconstruct the plasma boundary shapes as well as the eddy current profiles in the UTST (The University of Tokyo) and ST40 device (Tokamak Energy Ltd). The present research results demonstrate the effectiveness of the M-CCS method in the reconstruction analyses of ST merging.

  9. Identification of waves by RF magnetic probes during lower hybrid wave injection experiments on the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Shinya, Takahiro; Ejiri, Akira; Takase, Yuichi

    2014-01-01

    RF magnetic probes can be used to measure not only the wavevector, but also the polarization of waves in plasmas. A 5-channel RF magnetic probe (5ch-RFMP) was installed in the TST-2 spherical tokamak and the waves were studied in detail during lower hybrid wave injection experiments. From the polarization measurements, the poloidal RF magnetic field is found to be dominant. In addition to polarization, components of k perpendicular to the major radial direction were obtained from phase differences among the five channels. The radial wavenumber was obtained by scanning the radial position of the 5ch-RFMP on a shot by shot basis. The measured wavevector and polarization in the plasma edge region were consistent with those calculated from the wave equation for the slow wave branch. While the waves with small and large k ∥ were excited by the antenna, only the small k ∥ component was measured by the 5ch-RFMP; this suggests that the waves with larger k ∥ were absorbed by the plasma. (author)

  10. Divertor impurity injection using high voltage arcs for impurity transport studies on the Mega Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Leggate, H. J.; Turner, M. M.; Lisgo, S. W.; Harrison, J. R.; Elmore, S.; Allan, S. Y.; Gaffka, R. C.; Stephen, R. C.

    2014-01-01

    The operation of next-generation fusion reactors will be significantly affected by impurity transport in the scrape-off layer (SOL). Current modelling efforts are restricted by a lack of detailed data on impurity transport in the SOL. In order to address this, a carbon injector has been designed and installed on the Mega Amp Spherical Tokamak (MAST). The injector creates short lived carbon plumes originating at the MAST divertor lasting less than 50 μs. High voltage capacitor banks are used to create a discharge across concentric carbon electrodes located in a probe mounted on the Divertor Science Facility in the MAST lower divertor. This results in a very short plume duration allowing observation of the evolution of the plume and precise localisation of the plume relative to the X-point on MAST. The emission from the carbon plume was imaged using fast visible cameras filtered in order to isolate the carbon II and carbon III emission lines centered around 514 nm and 465 nm

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  12. Recent QUEST experiments on non-inductive current drive and plasma-wall interaction towards steady state operation of spherical tokamak

    International Nuclear Information System (INIS)

    Hanada, K.; Zushi, H.; Idei, H.; Nakamura, K.; Nagashima, Y.; Hasegawa, M.; Fujisawa, A.; Higashijima, A.; Kawasaki, S.; Nakashima, H.; Ishiguro, M.; Tashima, S.; Kalinnikova, E.I.; Mitarai, O.; Maekawa, T.; Fukuyama, A.; Takase, Y.; Gao, X.; Liu, H.; Qian, J.; Ono, M.; Raman, R.; Peng, M.

    2015-01-01

    Full text of publication follows. Steady state operation (SSO) of magnetic fusion devices is one of the goals for fusion research. Development of non-inductive current drive and investigation of plasma-wall interaction (PWI) are issues to be resolved for SSO. Because of the very limited central solenoid (CS) flux in a spherical tokamak (ST), methods for non-inductive plasma current start-up and sustainment are necessary. Fully non-inductive plasma up to approximately 5 min was successfully demonstrated on the spherical tokamak QUEST. Furthermore, recharging of the center solenoid coil was also achieved in OH+RF plasmas with plasma current feedback using the CS. During the plasma start-up phase, precession motion of trapped electrons can drive some current, which plays an essential role in forming a closed flux surface. On QUEST, the main parts of the plasma facing components (PFCs) are covered by tungsten plates (W) or coated by W plasma spray and are actively cooled by water circulation. The increase in water temperature quantitatively provides the deposited power to each PFC. The power balance during long duration discharges has been studied for various types of magnetic configurations such as limiter, upper and lower single-null divertor discharges. As, the temperature of any PFCs reaches a steady-state condition during long pulse, the power balance can be obtained. It is found that the discharge duration of QUEST is significantly limited by particle imbalance shown by gradual increment of plasma and neutral density. The additional influx of neutrals was provided by recycling of hydrogen, which is still uncontrollable. A point model of particle balance was applied to a long-duration divertor discharge, and it was found that a small increment of particle-influx occurred around the end of the long duration discharge. A post-mortem analysis of surface-attaching specimen during an experimental campaign indicates that the increased amount of neutral influx could be

  13. Development of Atomic Beam Probe for tokamaks

    Czech Academy of Sciences Publication Activity Database

    Berta, M.; Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S.; Havlíček, Josef; Háček, Pavel

    2013-01-01

    Roč. 88, č. 11 (2013), s. 2875-2880 ISSN 0920-3796 R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ABP * Plasma diagnostics * COMPASS tokamak * Current density * Plasma density profile measurement Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613005048#

  14. Development of a visualized software for tokamak experiment data processing

    International Nuclear Information System (INIS)

    Cao Jianyong; Ding Xuantong; Luo Cuiwen

    2004-01-01

    With the VBA programming in Microsoft Excel, the authors have developed a post-processing software of experimental data in tokamak. The standard formal data in the HL-1M and HL-2A tokamaks can be read, displayed in Excel, and transmitted directly into the MATLAB workspace, for displaying pictures in MATLAB with the software. The authors have also developed data post-processing software in MATLAB environment, which can read standard format data, display picture, supply visual graphical user interface and provide part of advanced signal processing ability

  15. Dependence of the fast waves-plasma interactions in pre-heated spherical tokamaks on the antenna location and poloidal extension

    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

  16. Assessment of power deposition dependence on the antenna poloidal extension in the fast waves-plasma interaction in pre-heated spherical tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Komoshvili, K [Tel Aviv University, Ramat Aviv (Israel); Cuperman, S [Tel Aviv University, Ramat Aviv (Israel); Bruma, C [Tel Aviv University, Ramat Aviv (Israel)

    2007-09-15

    To assess the effect of antenna poloidal extension on fast waves-plasma interactions in pre-heated spherical tokamaks and, as a result, to assist the determination of optimal conditions for power deposition, we carried out a global, numerical investigation. Thus, we solved the steady-state full wave equations for Alfvenic modes in an inhomogeneous, non-uniformly magnetized, resistive, low aspect ratio tokamak plasma with appropriate consideration of boundary conditions; in this, processes such as wave propagation, reflection, transmission, absorption and mode conversion as well as mode-coupling(s) by plasma cross-section non-homogeneity generated waves were included. The results were analysed in terms of the directions of the current densities generated in the presence of up low field side or down high field side magnetic field gradient. Suitable antenna location and poloidal extension for maximum power deposition were determined.

  17. Assessment of power deposition dependence on the antenna poloidal extension in the fast waves-plasma interaction in pre-heated spherical tokamaks

    International Nuclear Information System (INIS)

    Komoshvili, K; Cuperman, S; Bruma, C

    2007-01-01

    To assess the effect of antenna poloidal extension on fast waves-plasma interactions in pre-heated spherical tokamaks and, as a result, to assist the determination of optimal conditions for power deposition, we carried out a global, numerical investigation. Thus, we solved the steady-state full wave equations for Alfvenic modes in an inhomogeneous, non-uniformly magnetized, resistive, low aspect ratio tokamak plasma with appropriate consideration of boundary conditions; in this, processes such as wave propagation, reflection, transmission, absorption and mode conversion as well as mode-coupling(s) by plasma cross-section non-homogeneity generated waves were included. The results were analysed in terms of the directions of the current densities generated in the presence of up low field side or down high field side magnetic field gradient. Suitable antenna location and poloidal extension for maximum power deposition were determined

  18. Software development of the KSTAR Tokamak Monitoring System

    International Nuclear Information System (INIS)

    Kim, K.H.; Lee, T.G.; Baek, S.; Lee, S.I.; Chu, Y.; Kim, Y.O.; Kim, J.S.; Park, M.K.; Oh, Y.K.

    2008-01-01

    The Korea Superconducting Tokamak Advanced Research (KSTAR) project, which is constructing a superconducting Tokamak, was launched in 1996. Much progress in instrumentation and control has been made since then and the construction phase will be finished in August 2007. The Tokamak Monitoring System (TMS) measures the temperatures of the superconducting magnets, bus-lines, and structures and hence monitors the superconducting conditions during the operation of the KSTAR Tokamak. The TMS also measures the strains and displacements on the structures in order to monitor the mechanical safety. There are around 400 temperature sensors, more than 240 strain gauges, 10 displacement gauges and 10 Hall sensors. The TMS utilizes Cernox sensors for low temperature measurement and each sensor has its own characteristic curve. In addition, the TMS needs to perform complex arithmetic operations to convert the measurements into temperatures for each Cernox sensor for this large number of monitoring channels. A special software development effort was required to reduce the temperature conversion time and multi-threading to achieve the higher performance needed to handle the large number of channels. We have developed the TMS with PXI hardware and with EPICS software. We will describe the details of the implementations in this paper

  19. 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)

  20. Plasma current start-up experiments without the central solenoid in the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Takase, Y.; Ejiri, A.; Shiraiwa, S.; Adachi, Y.; Ishii, N.; Kasahara, H.; Nuga, H.; Ono, Y.; Oosako, T.; Sasaki, M.; Shimada, Y.; Sumitomo, N.; Taguchi, I.; Tojo, H.; Tsujimura, J.; Ushigome, M.; Yamada, T.; Hanada, K.; Hasegawa, M.; Idei, H.; Nakamura, K.; Sakamoto, M.; Sasaki, K.; Sato, K.N.; Zushi, H.; Nishino, N.; Mitarai, O.

    2006-01-01

    Several techniques for initiating the plasma current without the use of the central solenoid are being developed in TST-2. While TST-2 was temporarily located at Kyushu University, two types of start-up scenarios were demonstrated. (1) A plasma current of 4 kA was generated and sustained for 0.28 s by either electron cyclotron wave or electron Bernstein wave, without induction. (2) A plasma current of 10 kA was obtained transiently by induction using only outboard poloidal field coils. In the second scenario, it is important to supply sufficient power for ionization (100 kW of EC power was sufficient in this case), since the vertical field during start-up is not adequate to maintain plasma equilibrium. In addition, electron heating experiments using the X-B mode conversion scenario were performed, and a heating efficiency of 60% was observed at a 100 kW RF power level. TST-2 is now located at the Kashiwa Campus of the University of Tokyo. Significant upgrades were made in both magnetic coil power supplies and RF systems, and plasma experiments have restarted. RF power of up to 400 kW is available in the high-harmonic fast wave frequency range around 20 MHz. Four 200 MHz transmitters are now being prepared for plasma current start-up experiments using RF power in the lower-hybrid frequency range. Preparations are in progress for a new plasma merging experiment (UTST) aimed at the formation and sustainment of ultra-high β ST plasmas

  1. 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

  2. development of a hydrothermal method to synthesize spherical znse

    African Journals Online (AJOL)

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    nanoparticles have a zinc blend structure and in a spherical form with ... optoelectronic devices such as blue-green laser diodes and turnable mid-IR ... Solvothermal methods have also been developed for the synthesis of ZnSe and CdSe. The.

  3. Progress in Developing a High-Availability Advanced Tokamak Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, T.; Goldston, R.; Kessel, C.; Neilson, G.; Menard, J.; Prager, S.; Scott, S.; Titus, P.; Zarnstorff, M., E-mail: tbrown@pppl.gov [Princeton University, Princeton Plasma Physics Laboratory, Princeton (United States); Costley, A. [Henley on Thames (United Kingdom); El-Guebaly, L. [University of Wisconsin, Madison (United States); Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Waganer, L. [St. Louis (United States)

    2012-09-15

    Full text: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The mission of the pilot plant was set to encompass component test and fusion nuclear science missions yet produce net electricity with high availability in a device designed to be prototypical of the commercial device. The objective of the study was to evaluate three different magnetic configuration options, the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS) in an effort to establish component characteristics, maintenance features and the general arrangement of each candidate device. With the move to look beyond ITER the fusion community is now beginning to embark on DEMO reactor studies with an emphasis on defining configuration arrangements that can meet a high availability goal. In this paper the AT pilot plant design will be presented. The selected maintenance approach, the device arrangement and sizing of the in-vessel components and details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will be discussed. Efforts made to enhance the interaction of in-vessel maintenance activities, the hot cell and the transfer process to develop simplifying solutions will also be addressed. (author)

  4. Experimental demonstration of tokamak inductive flux saving by transient coaxial helicity injection on national spherical torus experiment

    Energy Technology Data Exchange (ETDEWEB)

    Raman, R.; Jarboe, T. R.; Nelson, B. A. [University of Washington, Seattle, Washington 98195 (United States); Mueller, D.; Bell, M. G.; Gerhardt, S.; LeBlanc, B.; Menard, J.; Ono, M.; Roquemore, L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2011-09-15

    Discharges initiated by transient coaxial helicity injection in National Spherical Torus Experiment have attained peak toroidal plasma currents up to 300 kA. When induction from the central solenoid is then applied, these discharges develop up to 300 kA additional current compared to discharges initiated by induction only. CHI initiated discharges in NSTX have achieved 1 MA of plasma current using only 258 mWb of solenoid flux whereas standard induction-only discharges require about 50% more solenoid flux to reach 1 MA. In addition, the CHI-initiated discharge has lower plasma density and a low normalized internal plasma inductance of 0.35, as needed for achieving advanced scenarios in NSTX.

  5. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  7. Structure and parameters dependences of Alfven wave current drive generated in the low-field side of simulated spherical tokamaks

    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)

  8. Structure and parameters dependences of Alfven wave current drive generated in the low-field side of simulated spherical tokamaks

    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)

  9. Development of large insulator rings for the TOKAMAK Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1977-01-01

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

  10. Analysis of Shield Construction in Spherical Weathered Granite Development Area

    Science.gov (United States)

    Cao, Quan; Li, Peigang; Gong, Shuhua

    2018-01-01

    The distribution of spherical weathered bodies (commonly known as "boulder") in the granite development area directly affects the shield construction of urban rail transit engineering. This paper is based on the case of shield construction of granite globular development area in Southern China area, the parameter control in shield machine selection and shield advancing during the shield tunneling in this special geological environment is analyzed. And it is suggested that shield machine should be selected for shield construction of granite spherical weathered zone. Driving speed, cutter torque, shield machine thrust, the amount of penetration and the speed of the cutter head of shield machine should be controlled when driving the boulder formation, in order to achieve smooth excavation and reduce the disturbance to the formation.

  11. The effect of off-axis neutral beam injection on sawtooth stability in ASDEX Upgrade and Mega-Ampere Spherical Tokamak

    International Nuclear Information System (INIS)

    Chapman, I. T.; de Bock, M. F.; Pinches, S. D.; Turnyanskiy, M. R.; Igochine, V. G.; Maraschek, M.; Tardini, G.

    2009-01-01

    Sawtooth behavior has been investigated in plasmas heated with off-axis neutral beam injection in ASDEX Upgrade [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)] and the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes et al., Nucl. Fusion 41, 1423 (2001)]. Provided that the fast ions are well confined, the sawtooth period is found to decrease as the neutral beam is injected further off-axis. Drift kinetic modeling of such discharges qualitatively shows that the passing fast ions born outside the q=1 rational surface can destabilize the n=1 internal kink mode, thought to be related to the sawtooth instability. This effect can be enhanced by optimizing the deposition of the off-axis beam energetic particle population with respect to the mode location.

  12. Development of a tokamak plasma optimized for stability and confinement

    International Nuclear Information System (INIS)

    Politzer, P.A.

    1995-02-01

    Design of an economically attractive tokamak fusion reactor depends on producing steady-state plasma operation with simultaneous high energy density (β) and high energy confinement (τ E ); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H≡ τ E /τ ITER-89P = 4) and high normalized β (β N ≡ β/(I/aB) = 6%-m-T/MA) have been obtained. For the present, these conditions have been produced separately and in transient discharges. The DIII-D advanced tokamak development program is directed toward developing an understanding of the characteristics which lead to high stability and confinement, and to use that understanding to demonstrate stationary, high performance operation through active control of the plasma shape and profiles. The authors have identified some of the features of the operating modes in DIII-D that contribute to better performance. These are control of the plasma shape, control of both bulk plasma rotation and shear in the rotation and Er profiles, and particularly control of the toroidal current profiles. In order to guide their future experiments, they are developing optimized scenarios based on their anticipated plasma control capabilities, particularly using fast wave current drive (on-axis) and electron cyclotron current drive (off-axis). The most highly developed model is the second-stable core VH-mode, which has a reversed magnetic shear safety factor profile [q(O) = 3.9, q min = 2.6, and q 95 = 6]. This model plasma uses profiles which the authors expect to be realizable. At β N ≥ 6, it is stable to n=l kink modes and ideal ballooning modes, and is expected to reach H ≥ 3 with VH-mode-like confinement

  13. Design and development of the helicity injection system in Versatile Experiment Spherical Torus

    International Nuclear Information System (INIS)

    Park, JongYoon; An, Younghwa; Jung, Bongki; Lee, Jeongwon; Lee, HyunYoung; Chung, Kyoung-Jae; Na, Yong-Su; Hwang, Y.S.

    2015-01-01

    Graphical abstract: - Highlights: • A high current electron gun with single pulse power for both arc and extraction is developed. • The optimal gun operation is confirmed by impedance matching between the PFN and plasma. • The gun injected currents of 0.95 kA with the voltage of ∼410 V for 5 ms with a 1.2 kV PFN. • The helicity injection system using the gun has been developed and tested successfully in VEST. • Toroidal currents of up to 3.8 kA confirm possible relaxation into tokamak-like plasma. - Abstract: A helicity injection system for the Versatile Experiment Spherical Torus (VEST) has been successfully developed and commissioned. A high current electron gun utilizing hollow cathode and washer stacks has been designed and constructed with a single pulse power system that can provide voltages for both arc discharge and extraction sequentially. Tests for electron gun operation with the single pulse power system have been conducted under various toroidal and poloidal field strengths. The estimated plasma impedance, depending on the injection magnetic field structure, can be utilized for the optimal gun operation by impedance matching between the pulse power system and plasma. With the charging voltage of 1.2 kV, injection current of 0.95 kA has been obtained with the injection voltage of 410 V for about 5 ms. Initial helicity injection experiments have been conducted under various toroidal and poloidal field strengths and a toroidal plasma current of up to 3.8 kA is observed with the current multiplication larger than the geometric stacking ratio, confirming the possibility of relaxation into tokamak-like plasma with closed flux formation.

  14. Design and development of the helicity injection system in Versatile Experiment Spherical Torus

    Energy Technology Data Exchange (ETDEWEB)

    Park, JongYoon; An, Younghwa; Jung, Bongki; Lee, Jeongwon; Lee, HyunYoung; Chung, Kyoung-Jae; Na, Yong-Su; Hwang, Y.S., E-mail: yhwang@snu.ac.kr

    2015-10-15

    Graphical abstract: - Highlights: • A high current electron gun with single pulse power for both arc and extraction is developed. • The optimal gun operation is confirmed by impedance matching between the PFN and plasma. • The gun injected currents of 0.95 kA with the voltage of ∼410 V for 5 ms with a 1.2 kV PFN. • The helicity injection system using the gun has been developed and tested successfully in VEST. • Toroidal currents of up to 3.8 kA confirm possible relaxation into tokamak-like plasma. - Abstract: A helicity injection system for the Versatile Experiment Spherical Torus (VEST) has been successfully developed and commissioned. A high current electron gun utilizing hollow cathode and washer stacks has been designed and constructed with a single pulse power system that can provide voltages for both arc discharge and extraction sequentially. Tests for electron gun operation with the single pulse power system have been conducted under various toroidal and poloidal field strengths. The estimated plasma impedance, depending on the injection magnetic field structure, can be utilized for the optimal gun operation by impedance matching between the pulse power system and plasma. With the charging voltage of 1.2 kV, injection current of 0.95 kA has been obtained with the injection voltage of 410 V for about 5 ms. Initial helicity injection experiments have been conducted under various toroidal and poloidal field strengths and a toroidal plasma current of up to 3.8 kA is observed with the current multiplication larger than the geometric stacking ratio, confirming the possibility of relaxation into tokamak-like plasma with closed flux formation.

  15. Recent developments in Bayesian inference of tokamak plasma equilibria and high-dimensional stochastic quadratures

    International Nuclear Information System (INIS)

    Von Nessi, G T; Hole, M J

    2014-01-01

    We present recent results and technical breakthroughs for the Bayesian inference of tokamak equilibria using force-balance as a prior constraint. Issues surrounding model parameter representation and posterior analysis are discussed and addressed. These points motivate the recent advancements embodied in the Bayesian Equilibrium Analysis and Simulation Tool (BEAST) software being presently utilized to study equilibria on the Mega-Ampere Spherical Tokamak (MAST) experiment in the UK (von Nessi et al 2012 J. Phys. A 46 185501). State-of-the-art results of using BEAST to study MAST equilibria are reviewed, with recent code advancements being systematically presented though out the manuscript. (paper)

  16. Control System Development Plan for the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Neumeyer, C.; Mueller, D.; Gates, D.A.; Ferron, J.R.

    1999-01-01

    The National Spherical Torus Experiment (NSTX) has as one of its primary goals the demonstration of the attractiveness of the spherical torus concept as a fusion power plant. Central to this goal is the achievement of high plasma β ( = 2 micro 0 /B 2 a measure of the efficiency of a magnetic plasma confinement system). It has been demonstrated both theoretically and experimentally that the maximum achievable β is a strong function of both local and global plasma parameters. It is therefore important to optimize control of the plasma. To this end a phased development plan for digital plasma control on NSTX is presented. The relative level of sophistication of the control system software and hardware will be increased according to the demands of the experimental program in a three phase plan. During Day 0 (first plasma), a simple coil current control algorithm will initiate plasma operations. During the second phase (Day 1) of plasma operations the control system will continue to use the preprogrammed algorithm to initiate plasma breakdown but will then change over to a rudimentary plasma control scheme based on linear combinations of measured plasma fields and fluxes. The third phase of NSTX plasma control system development will utilize the rtEFIT code, first used on DIII-D, to determine, in real-time, the full plasma equilibrium by inverting the Grad-Shafranov equation. The details of the development plan, including a description of the proposed hardware will be presented

  17. Development of remote control integrator system on Tokamak

    International Nuclear Information System (INIS)

    Wu Yichun; Wang Lingzhi; Shu Shuangbao

    2014-01-01

    In order to meet with the requirement of electromagnetic diagnosis to the J-TEXT Tokamak, a remote control integrator system was developed. With modular design method, the integrator system is composed of the integrator cards, a control card, a linear power card and the BNC interface cards, and it uses the PC control soft- ware to conduct network control. An integrator system provides 32 integrator channels, and all integral channels have four kinds of integral time constants for remote selection and provide three kinds of integrator running control methods. According to laboratory and J-TEXT field testing, it shows that the output voltage range is -10-10 V, output noise is not more than 5 mV, and for the four kinds of integral time constants, the integral output drifts are all less than 5 mV within 100 s for each integrator channel. (authors)

  18. X-ray measurements during plasma current start-up experiments using the lower hybrid wave on the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Wakatsuki, Takuma; Ejiri, Akira; Kakuda, Hidetoshi

    2012-01-01

    Non-inductive plasma current start-up experiments using RF power in the lower hybrid frequency range is being conducted on the TST-2 spherical tokamak. Plasma currents of up to 15 kA have been achieved. The effect of direct current drive can be seen by comparing the cases with co-drive and counter-drive. X-rays in various energy ranges were measured to investigate the interaction between the wave and the electrons. Soft X-ray (SX) measurements revealed that the perpendicular SX emission increased significantly as the plasma current increased, and that the tangential SX emission in the direction of RF drive was enhanced more strongly in the co-drive case compared to the counter-drive case. These observations imply that the fast electrons accelerated by the lower hybrid wave contribute to the plasma current. However, RF amplitude modulation experiments showed that the confinement time of these fast electrons are very short (less than 0.05 ms), much shorter than the collisional slowing down time. Hard X-ray spectral measurements showed that the radiation temperature of fast electrons in the co-direction for current drive was higher than that in the counter-direction. These observations are consistent with the existence of RF-driven fast electrons. (author)

  19. Active particle control experiments and critical particle flux discriminating between the wall pumping and fuelling in the compact plasma wall interaction device CPD spherical tokamak

    International Nuclear Information System (INIS)

    Zushi, H.; Sakamoto, M.; Yoshinaga, T.; Higashizono, Y.; Hanada, K.; Yoshida, N.; Tokunaga, K.; Kawasaki, S.; Sato, K. N.; Nakamura, K.; Idei, H.; Hirooka, Y.; Bhattacharyay, R.; Okamoto, K.; Miyazaki, T.; Honma, H.; Nakashima, Y.; Nishino, N.; Kado, S.; Shikama, T.

    2009-01-01

    Two approaches associated with wall recycling have been performed in a small spherical tokamak device CPD (compact plasma wall interaction experimental device), that is, (1) demonstration of active particle recycling control, namely, 'active wall pumping' using a rotating poloidal limiter whose surface is continuously gettered by lithium and (2) a basic study of the key parameters which discriminates between 'wall pumping and fuelling'. For the former, active control of 'wall pumping' has been demonstrated during 50 kW RF current drive discharges whose pulse length is typically ∼300 ms. Although the rotating limiter is located at the outer board, as soon as the rotating drum is gettered with lithium, hydrogen recycling measured with H α spectroscopy decreases by about a factor of 3 not only near the limiter but also in the centre stack region. Also, the oxygen impurity level measured with O II spectroscopy is reduced by about a factor of 3. As a consequence of the reduced recycling and impurity level, RF driven current has nearly doubled at the same vertical magnetic field. For the latter, global plasma wall interaction with plasma facing components in the vessel is studied in a simple torus produced by electron cyclotron waves with I p -4 to ∼0.1 x 10 -4 Torr during the experimental campaign (∼3000 shots). In the wall pumping pressure range the wall pumping fraction is reduced with increasing surface temperature up to 150 deg. C.

  20. Advanced fusion technologies developed for JT-60 superconducting tokamak

    International Nuclear Information System (INIS)

    Sakasai, Akira; Ishida, S.; Matsukawa, M.

    2003-01-01

    The modification of JT-60U is planned as a full superconducting tokamak (JT-60SC). The objectives of the JT-60SC program are to establish scientific and technological bases for the steady-state operation of high performance plasmas and utilization of reduced-activation materials in economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to DEMO reactor have been developed in the superconducting magnet technology and plasma facing components for the design of JT-60SC. To achieve a high current density in a superconducting strand, Nb 3 Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFC) of JT-60SC. The R and D to demonstrate applicability of Nb 3 Al conductor to the TFC by a react-and-wind technique have been carried out using a full-size Nb 3 Al conductor. A full-size NbTi conductor with low AC loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the CFC target was successfully demonstrated on the electron beam irradiation stand. (author)

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

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1978-01-01

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

  2. Economic considerations of commercial tokamak options

    International Nuclear Information System (INIS)

    Dabiri, A.E.

    1986-05-01

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

  3. Development of lab scale fast gas injection system for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Pathan, F.S.; Banaudha, Moni; Khristi, Yohan; Khan, M.S.; Khan, Ziauddin; Raval, D.C.; Khirwadkar, Samir

    2017-01-01

    The plasma density control plays an important role in Tokamak operation. The factors that influence plasma density in a Tokamak device are working gas injection, pumping, ionization rate and the recycle coefficient representing the wall conditions. Among these factors, gas injection is relatively convenient to be controlled. Hence, the most frequently adopted method to control the plasma density is to control the fast gas injection. This paper describes the design and experimental work carried out towards the development of Fast Gas Injection System for SST-1 Tokamak. Laboratory based test setup was successfully established for Fast Gas Injection System that can feed predefined quantity of gas in a controlled manner into vacuum chamber. Further, this FGIS system will be implemented in SST-1 Tokamak environment with online density feedback signal

  4. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

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

  5. Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, California 90095 (United States); Garofalo, A M [Columbia University, New York, New York 10027 (United States); Greenfield, C M [General Atomics, San Diego, California 92186-5608 (United States); Kaye, S M [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Menard, J E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Murakami, M [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sabbagh, S A [Columbia University, New York, New York 10027 (United States); Austin, M E [University of Texas-Austin, Austin, Texas 78712 (United States); Bell, R E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Burrell, K H [General Atomics, San Diego, California 92186-5608 (United States); Ferron, J R [General Atomics, San Diego, California 92186-5608 (United States); Gates, D A [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Groebner, R J; Hyatt, A W; Luce, T C; Petty, C C; Wade, M R; Waltz, R E [General Atomics, San Diego, California 92186-5608 (United States); Jayakumar, R J [Lawrence Livermore National Lab., Livermore, California 94550 (United States); Kinsey, J E [Lehigh Univ., Bethlehem, Pennsylvania 18015 (United States); LeBlanc, B P [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); McKee, G R [Univ. of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Okabayashi, M [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Peng, Y-K M [Oak Ridge National Lab., Oak Ridge, Tennessee 37831 (United States); Politzer, P A [General Atomics, San Diego, California 92186-5608 (United States); Rhodes, T L [Dept. of Electrical Engineering and PSTI, Univ. of California, Los Angeles, California 90095 (United States)

    2006-12-15

    Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f{sub BS} {>=} 60%, q{sub 95} {approx} 4-5 and G {identical_to}{beta}{sub N}H{sub scaling}/q{sub 95}{sup 2} {>=}0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E ? B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at {beta}{sub N} {>=} 4 with q{sub min} {>=} 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. {beta}{sub N} = 4, H{sub 89} = 2.5, with f{sub BS} > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q {>=} 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

  6. 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

  7. Development of Tokamak experiment technology - Study of ICRF coupling in the KAIST tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Duk In; Chang, Hang Young; Lee, Soon Chil; Kwon, Gi Chung; Seo, Sung Hun; Jeon, Sang Jin; Heo, Sung Hee; Heo, Eun Gi; Lee, Dae Hang; Lee, Chan Hee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-08-01

    Research objectives are to design and fabricate antenna, measure the property of absorption transmitted to the plasma, and research the physical phenomena about the ICRF coupling. Main heating method is ohmic heating at the KAIST tokamak. So, the plasma current produced is more than 30 kA and, the loop voltage of the plasma is 2 {approx} 3V. The power of the plasma by ohmic heating is about 100 kW. Because the toroidal field is 5 {approx} 8 kG, it is needed RF system with more than 100 kW in 7 {approx} 15 MHz. In the first year a RF amplifier with 1 kW in 300 khz {approx} 35 MHz was bought. The manufacture of ICRF system will start from next years. In the research on antenna, we study the method how to measure electric field emitted from antenna using piezo elements. Experimentally, we obtain the results that the signal of piezo element is proportional to the square of electric field. In the next year, we will research the type of antenna subsequently. 28 refs., 3 tabs., 18 figs. (author)

  8. Development on JET of Advanced Tokamak Operations for ITER

    International Nuclear Information System (INIS)

    Tuccillo, A.A.; Crisanti, F.; Litaudon, X.

    2005-01-01

    Recent research on Advanced Tokamak in JET has focused on scenarii with both monotonic and reversed shear q profiles having plasma parameters as relevant as possible for extrapolation to ITER. Wide ITBs, R∼3.7m, are formed at ITER relevant triangularity δ∼0.45, with n e /n G ∼60% and ELMs moderated by Ne injection. At higher current (I P ≤3.5MA, δ∼0.25) wide ITBs sitting at R≥ 3.5m (positive shear region) have been developed, generally MHD events terminate these barrier otherwise limited in strength by power availability. ITBs with core density close to Greenwald value are obtained with plasma target preformed by opportune timing of LHCD, pellet injection and small amount of NBI power. ITB start with toroidal rotation 4 times lower than the standard NBI heated ITBs. Full CD is achieved in reversed shear ITBs at 3T/1.8 MA, by using 10MW NBI, 5MW ICRH and 3MW LH. Wide ITBs located at R=3.6m, without impurity accumulation and type-III ELMs edge can be sustained for a time close to neo-classical resistive time. These discharges have been extended to the maximum duration allowed by subsystems (20s) with the JET record of injected energy: E∼330 MJ. Integrated control of pressure and current profile isit; feature used in these discharges. Central ICRF mode conversion electron heating, added to about 14MW NBI power, produced impressive ITBs with equivalent Q DT ∼ 0.25. Conversely ion ITBs are obtained with low torque injection, by ICRH 3 He minority heating of ions, on pure LHCD electron ITBs. Similarity experiments between JET and AUG have compared the dynamics of ITBs and have been the starting point of Hybrid Scenarios activity, then developed at ρ* as low as ρ*∼3*10 -3 . The development of hybrid regime with dominant electron heating has also started. Injection of trace of tritium and a mixture of Ar/Ne allowed studying fuel and impurities transport in many of the explored AT scenarios. (author)

  9. The spherical harmonics method, 1 (general development of the theory)

    International Nuclear Information System (INIS)

    Mark, C.

    1957-02-01

    A method of obtaining approximate solutions of the transport equation is presented in a form applicable in principle to any geometry. The approximation will give good results in cases where the angular distribution is not very anisotropic. The basis of the approximation is to expand the density per unit solid angle Ψ(→/r, →/Ω) in spherical harmonic tensors formed from →/Ω the unit vector in the direction of velocity, and to break off the expansion. A differential equation whose degree increases with the order of the approximation is obtained for the total density Ψ (o) (r). This equation has the form where the numbers ν i depend on the order of the approximation and on the value of the parameter a of the medium, but not at all on the geometry. When the equation for the total density is an ordinary equation, we simulate the physical condition of continuity of Ψ(→/r, →/Ω) at a boundary in a multi-medium problem by requiring that the spherical harmonic moments of Ψ(→/r, →/Ω) which we retain be continuous; and this determines the constants in the solution for Ψ (o) (→/r. The form of the solution for the total density and the necessary moments in an approximation of general order is given explicitly for plane and spherical symmetry; and for cylindrical symmetry the solution is given for two low-order approximations. In a later report (CRT-338, Revised) the application of the method to several problems involving plane and spherical symmetry will be discussed in detail and the results of a number of examples already worked will also be given. (author)

  10. The spherical harmonics method, 1 (general development of the theory)

    Energy Technology Data Exchange (ETDEWEB)

    Mark, C

    1957-02-15

    A method of obtaining approximate solutions of the transport equation is presented in a form applicable in principle to any geometry. The approximation will give good results in cases where the angular distribution is not very anisotropic. The basis of the approximation is to expand the density per unit solid angle {Psi}({yields}/r, {yields}/{Omega}) in spherical harmonic tensors formed from {yields}/{Omega} the unit vector in the direction of velocity, and to break off the expansion. A differential equation whose degree increases with the order of the approximation is obtained for the total density {Psi}{sup (o)}(r). This equation has the form where the numbers {nu}{sub i} depend on the order of the approximation and on the value of the parameter a of the medium, but not at all on the geometry. When the equation for the total density is an ordinary equation, we simulate the physical condition of continuity of {Psi}({yields}/r, {yields}/{Omega}) at a boundary in a multi-medium problem by requiring that the spherical harmonic moments of {Psi}({yields}/r, {yields}/{Omega}) which we retain be continuous; and this determines the constants in the solution for {Psi}{sup (o)}({yields}/r. The form of the solution for the total density and the necessary moments in an approximation of general order is given explicitly for plane and spherical symmetry; and for cylindrical symmetry the solution is given for two low-order approximations. In a later report (CRT-338, Revised) the application of the method to several problems involving plane and spherical symmetry will be discussed in detail and the results of a number of examples already worked will also be given. (author)

  11. History of the T-10 tokamak: Creation and development

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    2001-01-01

    The heating and thermal insulation of a hot plasma with the purpose of achieving a controlled thermonuclear reaction have been investigated for almost 50 years. Experiments in the T-10 tokamak, which have been carried out for 25 years, have played an important role in such investigations. This paper presents a history of the device, the physical and technological foundations underlying the project, and the results obtained in the ohmic heating mode during the first years of the device operation

  12. Immersive virtual walk-through development for tokamak using active head mounted display

    International Nuclear Information System (INIS)

    Dutta, Pramit

    2015-01-01

    A fully immersive virtual walk-through of the SST-1 tokamak has been developed. The virtual walkthrough renders the virtual model of SST-1 tokamak through a active stereoscopic head mounted display to visualize the virtual environment. All locations inside and outside of the reactor can be accessed and reviewed. Such a virtual walkthrough provides a 1:1 scale visualization of all components of the tokamak. To achieve such a virtual model, the graphical details of the tokamak CAD model are enhanced. Such enhancements are provided to improve lighting conditions at various locations, texturing of components to have a realistic visual effect and 360° rendering for ease of access. The graphical enhancements also include the redefinition of the facets to optimize the surface triangles to remove lags in display during visual rendering. Two separate algorithms are developed to interact with the virtual model. A fly-by algorithm, developed using C#, uses inputs from a commercial joystick to navigate within the virtual environment. The second algorithm uses the IR and gyroscopic tracking system of the head mounted display to render view as per the current pose of the user within the virtual environment and the direction of view. Such a virtual walk-thorough can be used extensively for design review and integration, review of new components, operator training for remote handling, operations, upgrades of tokamak, etc. (author)

  13. Development of superconducting poloidal field coils for medium and large size tokamaks

    International Nuclear Information System (INIS)

    Dittrich, H.-G.; Forster, S.; Hofmann, A.

    1983-01-01

    Large long pulse tokamak fusion experiments require the use of superconducting poloidal field (PF) coils. In the past not much attention has been paid to the development of such coils. Therefore a development programme has been initiated recently at KfK. In this report start with summarizing the relevant PF coil parameters of some medium and large size tokamaks presently under construction or design, respectively. The most important areas of research and development work are deduced from these parameters. Design considerations and first experimental results concerning low loss conductors, cooling concepts and structural components are given

  14. Scenario development during commissioning operations on the National Spherical Torus Experiment Upgrade

    Science.gov (United States)

    Battaglia, D. J.; Boyer, M. D.; Gerhardt, S.; Mueller, D.; Myers, C. E.; Guttenfelder, W.; Menard, J. E.; Sabbagh, S. A.; Scotti, F.; Bedoya, F.; Bell, R. E.; Berkery, J. W.; Diallo, A.; Ferraro, N.; Kaye, S. M.; Jaworski, M. A.; LeBlanc, B. P.; Ono, M.; Park, J.-K.; Podesta, M.; Raman, R.; Soukhanovskii, V.; NSTX-U Research, the; Operations; Engineering Team

    2018-04-01

    The National Spherical Torus Experiment Upgrade (NSTX-U) will advance the physics basis required for achieving steady-state, high-beta, and high-confinement conditions in a tokamak by accessing high toroidal fields (1 T) and plasma currents (1.0-2.0 MA) in a low aspect ratio geometry (A  =  1.6-1.8) with flexible auxiliary heating systems (12 MW NBI, 6 MW HHFW). This paper describes the progress in the development of L- and H-mode discharge scenarios and the commissioning of operational tools in the first ten weeks of operation that enable the scientific mission of NSTX-U. Vacuum field calculations completed prior to operations supported the rapid development and optimization of inductive breakdown at different values of ohmic solenoid current. The toroidal magnetic field (B T0  =  0.65 T) exceeded the maximum values achieved on NSTX and novel long-pulse L-mode discharges with regular sawtooth activity exceeded the longest pulses produced on NSTX (t pulse  >  1.8 s). The increased flux of the central solenoid facilitated the development of stationary L-mode discharges over a range of density and plasma current (I p). H-mode discharges achieved similar levels of stored energy, confinement (H98y,2  >  1) and stability (β N/β N-nowall  >  1) compared to NSTX discharges for I p  ⩽  1 MA. High-performance H-mode scenarios require an L-H transition early in the I p ramp-up phase in order to obtain low internal inductance (l i) throughout the discharge, which is conducive to maintaining vertical stability at high elongation (κ  >  2.2) and achieving long periods of MHD quiescent operations. The rapid progress in developing L- and H-mode scenarios in support of the scientific program was enabled by advances in real-time plasma control, efficient error field identification and correction, effective conditioning of the graphite wall and excellent diagnostic availability.

  15. Development of high thermal flux components for continuous operation in Tokamaks

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Coston, J.F.; Deschamps, P.; Lipa, M.

    1991-01-01

    High heat flux plasma facing components are under development and appropriate experimental evaluations have been carried out in order to operate during cycles of several hundred seconds. In Tore Supra, a large tokamak with a plasma nominal duration in excess of 30 seconds, solutions are tested that could be later applied to the NET/ITER tokamak, where peaked heat flux values of 15 MW/m 2 on the divertor plates are foreseen. The proposed concept is a swirl square tube design protected with brazed CFC flat tiles. Development programs and validation tests are presented. The tests results are compared with calculations

  16. The development of 128 ch fast multi channel pulse height analyzer for a tokamak plasmas

    International Nuclear Information System (INIS)

    Kawashima, Hisato; Matoba, Tohru; Ogawa, Toshihide; Kawakami, Tomohide

    1985-02-01

    A high counting rate multi channel pulse height analyzer was developed and tested to measure the detailed time evolution of X-ray energy spectrun radiated from a tokamak plasmas. Main developing objects of this analyzer are as follows. 1. The maximum counting rate and the minimum time resolution are 4 Mcps and 10 ms, respectively. 2. The energy resolution has ability to distinguish the characterisitic X-ray line. 3. Computer has to be used for operating system. This fast multi channel analyzer is using to measure the Soft X-ray spectrum on JFT-2M tokamak, and is confirmed to be useful for a practical measuring system. (author)

  17. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

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

  18. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

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

  19. Development and validation of a tokamak skin effect transformer model

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  20. Virtual reality applications in remote handling development for tokamaks in India

    International Nuclear Information System (INIS)

    Dutta, Pramit; Rastogi, Naveen; Gotewal, Krishan Kumar

    2017-01-01

    Highlights: • Evaluation of Virtual Reality (VR) in design and operation phases of Remote Handling (RH) equipment for tokamak. • VR based centralized facility, to cater RH development and operation, is setup at Institute for Plasma Research, India. • The VR facility system architecture and components are discussed. • Introduction to various VR applications developed for design and development of tokamak RH equipment. - Abstract: A tokamak is a plasma confinement device that can be used to achieve magnetically confined nuclear fusion within a reactor. Owing to the harsh environment, Remote Handling (RH) systems are used for inspection and maintenance of the tokamak in-vessel components. As the number of in-vessel components requiring RH maintenance is large, physical prototyping of all strategies becomes a major challenge. The operation of RH systems poses further challenge as all equipment have to be controlled remotely within very strict accuracy limits with minimum reliance on the available camera feedback. In both design and operation phases of RH equipment, application of Virtual Reality (VR) becomes imperative. The scope of this paper is to introduce some applications of VR in the design and operation cycle of RH, which are not available commercially. The paper discusses the requirement of VR as a tool for RH equipment design and operation. The details of a comprehensive VR facility that has been established to support the RH development for Indian tokamaks are also presented. Further, various cases studies are provided to highlight the utilization of this VR facility within phases of RH development and operation.

  1. Virtual reality applications in remote handling development for tokamaks in India

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, Pramit, E-mail: pramitd@ipr.res.in; Rastogi, Naveen; Gotewal, Krishan Kumar

    2017-05-15

    Highlights: • Evaluation of Virtual Reality (VR) in design and operation phases of Remote Handling (RH) equipment for tokamak. • VR based centralized facility, to cater RH development and operation, is setup at Institute for Plasma Research, India. • The VR facility system architecture and components are discussed. • Introduction to various VR applications developed for design and development of tokamak RH equipment. - Abstract: A tokamak is a plasma confinement device that can be used to achieve magnetically confined nuclear fusion within a reactor. Owing to the harsh environment, Remote Handling (RH) systems are used for inspection and maintenance of the tokamak in-vessel components. As the number of in-vessel components requiring RH maintenance is large, physical prototyping of all strategies becomes a major challenge. The operation of RH systems poses further challenge as all equipment have to be controlled remotely within very strict accuracy limits with minimum reliance on the available camera feedback. In both design and operation phases of RH equipment, application of Virtual Reality (VR) becomes imperative. The scope of this paper is to introduce some applications of VR in the design and operation cycle of RH, which are not available commercially. The paper discusses the requirement of VR as a tool for RH equipment design and operation. The details of a comprehensive VR facility that has been established to support the RH development for Indian tokamaks are also presented. Further, various cases studies are provided to highlight the utilization of this VR facility within phases of RH development and operation.

  2. 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

  3. Development of a measurement technique to characterize erosion and redeposition in a tokamak by speckle interferometry

    International Nuclear Information System (INIS)

    Dore, P.

    2006-11-01

    This work aims at proving the feasibility of temporal phase shifting speckle interferometry to make erosion/redeposition measurements on plasma facing components in situ on a tokamak. Results show clearly that the interferometric technique can be implemented on a tokamak to provide erosion/redeposition measurements. The optical setup and the interferograms acquisition and processing have been developed and tested in laboratory before being suited to the complex tokamak environment. We finally have an optical technique able to characterize erosion/redeposition mechanisms (amount of eroded/redeposited material, location) on optically rough plasma facing components (carbon fibre composite, tungsten). These components, suffering from random displacements (as vibrations) during acquisition, are relatively large (∼ 50 x 50 cm 2 ) and could be situated far away from the CCD camera (∼ 3 m). Now, we need to define the regions of plasma facing components where we want to make erosion and redeposition measurements. After that, we propose a diagnostic to validate the optical technique in situ on a tokamak, allowing us to develop a diagnostic for ITER. (author)

  4. Spherical sampling

    CERN Document Server

    Freeden, Willi; Schreiner, Michael

    2018-01-01

    This book presents, in a consistent and unified overview, results and developments in the field of today´s spherical sampling, particularly arising in mathematical geosciences. Although the book often refers to original contributions, the authors made them accessible to (graduate) students and scientists not only from mathematics but also from geosciences and geoengineering. Building a library of topics in spherical sampling theory it shows how advances in this theory lead to new discoveries in mathematical, geodetic, geophysical as well as other scientific branches like neuro-medicine. A must-to-read for everybody working in the area of spherical sampling.

  5. Development of the 'JFT-2' tokamak plasma position control system

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.

    1980-01-01

    Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)

  6. Development of LHCD launcher for next stage tokamak

    International Nuclear Information System (INIS)

    Seki, M.; Obara, K.; Maebara, S.

    1994-01-01

    In next stage LHCD experiment, long pulse RF injection is required for studying quasi-steady state tokamaks. The suppression of outgassing from waveguides is one of the main issues for LHCD launchers to transmit RF power in the waveguides continuously and stably. In order to know the parameters which control outgassing rate and to investigate how to reduce outgassing rate, JAERI and CEA have performed outgassing experiment by using four divided waveguides. The experimental setup and the results are reported. Steady state outgassing was observed in long duration up to 1800s when RF heat was removed by water cooling. In next generation LHCD launchers, it should be demanded to launch the high directive and sharp spectra, and to make the structure simple and compact. But these spectra require many waveguides in front of plasma, and this situation is not compatible with the compact structure which is necessary for low cost and easy maintenance. Moreover, the launchers are advantageous if the controllability is wide, and the low RF power density at grill mouth makes power launching easy. In order to realize the above features, a new launcher was devised. The conceptual structure is shown. The main R and D item is to divide RF power into three waveguides lined in poloidal direction. The RF property is discussed. (K.I.)

  7. Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

    International Nuclear Information System (INIS)

    Giunta, A.S.; Henderson, S.; O'Mullane, M.; Summers, H.P.; Harrison, J.; Doyle, J.G.

    2016-01-01

    This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

  8. Recent developments in engineering and technology concepts for prospective tokamak fusion reactors

    International Nuclear Information System (INIS)

    Ford, G.W.K.

    1987-01-01

    The tokamak has become the most developed magnetic fusion system and it appears likely that break-even and possibly ignition will first be demonstrated in existing machines of this type. Yet larger tokamaks could also demonstrate the essential technologies for the production of useful power. World-wide, well over a hundred tritium-breeder/heat-removal blanket concepts have been devised and preliminary engineering design studies undertaken, but the effort deployed on breeding and power recovery systems has been very small compared with that assigned to plasma research and development. The European Communities' NET (Next European Torus) project may offer an opportunity to redress this imbalance. The NET pre-design stage now in progress for some three years has selected many of the best features of plasma and nuclear design from the world's total efforts in these fields, and the NET concept is described in this paper as exemplifying where magnetic fusion power reactor technology stands today. It is concluded that although there are numerous more advanced types of magnetic confinement fusion reactor at early stages of their physics development, the tokamak offers the best opportunity for the early demonstration of fusion power

  9. KDAS: General-Purpose Data Acquisition System Developed for KAIST-Tokamak

    International Nuclear Information System (INIS)

    Seo, Seong-Heon; Choe, Wonho; Chang, Hong-Young; Jeong, Seung-Ho

    2000-01-01

    The Korea Advanced Institute of Science and Technology (KAIST)-Tokamak Data Acquisition System (KDAS) was originally developed for KAIST-Tokamak (R/a = 0.53 m/0.14 m). It operates on a distributed system based on personal computers and has a driver-based hierarchical structure. Since KDAS can be dynamically composed of any number of available computers, and the hardware-dependent codes can be thoroughly separated into external drivers, it exhibits excellent system performance flexibility and extensibility and can optimize various user needs. It collectively controls the VXI, CAMAC, GPIB, and RS232 instrument hybrids. With these useful and convenient features, it can be applied to any computerized experiment, especially to fusion-related research. The system design and features are discussed in detail

  10. Development of a remotely maintainable radio-frequency module for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Snider, J.D.

    1988-01-01

    The Compact Ignition Tokamak (CIT) will require reliable remote handling (RH) systems to overcome failures in diagnostic and operational equipment. Oak Ridge National laboratory (ORNL) is responsible for the ex-vessel remote maintenance systems for the CIT. Part of this effort is performing remote maintenance demonstrations on replicas of various CIT equipment. To ensure successful RH, the machine must be designed with proven remote maintenance features. In the demonstrations, critical remote maintenance features are tested before actual CIT equipment designs are finalized. Designs and procedures required to remotely remove and install a radio-frequency (rf) module from a modplane port on the tokamak were recently demonstrated at ORNL. This testing identified both successful design features for remote maintenance of the rf module and areas that require further development. 1 ref., 11 figs

  11. A study on the Fusion Reactor - Development of charge exchange recombination spectroscopy for tokamak diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tong Nyong; Kim, Dong Eon; Kim, Dae Sung; Kim, Seong Ho [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    1996-09-01

    This project has been carried to train people and accumulate the knowledge and techniques related to the measurement of the profiles of ion temperature, toroidal rotation velocity, and fully-stripped ion density in a fusion tokamak plasma by the development of plasma diagnostics using charge exchange recombination (CER) spectroscopy. Daring the 1 st year, the basic study and review on the charge exchange process and the conceptual design and review of the diagnostics have been conducted. In addition, the various atomic data centers around the world have been surveyed and atomic data related to CER have been constructed. The results of this project can be used to the construction and tokamak machine installation of a CER plasma diagnostic to a new superconducting supported by National Fusion Program. 42 refs., 3 tabs., 16 figs. (author)

  12. Development of DEMO-FNS tokamak for fusion and hybrid technologies

    Science.gov (United States)

    Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-07-01

    The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

  13. RECENT DEVELOPMENTS ON THE 110 GHz ELECTRON CYCLOTRON INSTATLLATION ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PONCE, D.; CALLIS, R.W.; CARY, W.P.; FERRON, J.R.; GREEN, M.; GRUNLOH, H.J.; GORELOV, Y.; LOHR, J.; ELLIS, R.A.

    2002-01-01

    OAK A271 RECENT DEVELOPMENTS ON THE 110 GHZ ELECTRON CYCLOTRON INSTALLATION ON THE DIII-D TOKAMAK. Significant improvements are being implement4ed to the capability of the 110 GHz electron cyclotron system on the DIII-D tokamak. Chief among these is the addition of the fifth and sixth 1 MW class gyrotrons, increasing the power available for auxiliary heating and current drive by nearly 60%. These tubes use artificially grown diamond rf output windows to obtain high power with long pulse capability. The beams from these tubes are nearly Gaussian, facilitating coupling to the waveguide. A new fully articulating dual launcher capable of high speed spatial scanning has been designed and tested. The launcher has two axis independent steering for each waveguide. the mirrors can be rotated at up to 100 o /s. A new feedback system linking the DIII-D Plasma Control System (PCS) with the gyrotron beam voltage waveform generators permits real-time feedback control of some plasma properties such as electron temperature. The PCS can use a variety of plasma monitors to generate its control signal, including electron cyclotron emission and Mirnov probes. Electron cyclotron heating and electron cyclotron current drive (ECH and ECCD) were used during this year's DIII-D experimental campaign to control electron temperature, density, and q profiles, induce an ELM-free H-mode, and suppress the m=2/n=1 neoclassical tearing mode. The new capabilities have expanded the role of EC systems in tokamak plasma control

  14. The dynamics of locked mode development in the T-11M tokamak

    International Nuclear Information System (INIS)

    Belov, A.M.

    2002-01-01

    Recent results of locked mode (LM) development studies in the T-11M tokamak are submitted. A particular interest in this type of plasma MHD-activity arises from the circumstance that an appropriate plasma perturbation is quasi-stationary and potentially could destroy plasma confinement, if it exceeds the same critical level. There are evidences to believe that LM amplitude approaches this critical level in the stage preceding a major disruption, resulting in the reduction of the magnetic shear in the plasma center, which finally initiates the disruption. (author)

  15. Technical development and operation of TV thomson scattering system on JFT-2M tokamak

    International Nuclear Information System (INIS)

    Shiina, Tomio; Yamauchi, Toshihiko; Ishige, Yoichi

    1998-10-01

    Six years have passed since the TV Thomson scattering system (TVTS) was completed and the operation was started on the JFT-2M tokamak. TVTS was developed in collaboration with Princeton Plasma Physics Laboratory. Many troubles on the hardware are and the software are were encountered. Improvements of the system were needed in each occasion. Phenomena of troubles were carefully analyzed and they have been solved in operating the system. This paper presents thus obtained know-how necessary for the operation of TVTS as well as methods of operation. (author)

  16. Multivariable shape control development on the DIII-D tokamak

    International Nuclear Information System (INIS)

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

    1997-11-01

    In this paper, the authors describe recent work on plasma shape and position control at DIII-D. This control consists of two equally challenging problems--the problem of identifying what the plasma actually looks like in real time, i.e. measuring the parameters to be controlled, and the task of determining the feedback algorithm which best controls these plasma parameters in a multiple input-output system. Recent implementation of the EFIT plasma equilibrium reconstruction algorithm in real time code which produces a new equilibrium estimate every 1.5 ms seems to solve the longstanding problem of obtaining sufficiently accurate plasma shape and position estimation. Stabilization of the open-loop unstable vertical motion is also viewed as a solved problem. The primary remaining problem appears to be how best to command the power supplies to achieve a desired shaping control response. They will describe the effort to understand and apply linearized models of plasma evolution to development and implementation of multivariable plasma controllers

  17. 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)

  18. Design and development of AXUV-based soft X-ray diagnostic camera for Aditya Tokamak

    International Nuclear Information System (INIS)

    Raval, Jayesh V.; Purohit, Shishir; Joisa, Y. Shankara

    2015-01-01

    The hot tokamak plasma emits Soft X-rays (SXR) in accordance with the temperature and density which are important to be studied. A silicon photo diode array (AXUV16ELG, Opto-diode, USA) based prototype SXR diagnostics is designed and developed for ADITYA tokamak for the study of SXR radial intensity profile, internal disruption (Saw-tooth crash), MHD instabilities. The diagnostic is having an array of 16 detector of millimeter dimension in a linear configuration. Absolute Extreme Ultra Violate (AXUV) detector offers compact size, improved time response with considerably good quantum efficiency in the soft X-ray range (200 eV to 10 keV). The diagnostic is designed in competence with the ADITYA tokamak protocol. The diagnostic design geometry allows detector view the plasma through a slot hole (0.5 cm X 0.05 cm), 10 μm Beryllium foil filter window, cutting off energies below 750 eV. The diagnostic was installed on Aditya vacuum vessel at radial port no 7 enabling the diagnostics to view the core plasma. The spatial resolution designed for diagnostic configuration is 1.3 cm at plasma centre. The signal generated from SXR detector is acquired with a dedicated single board computer based data acquisition system at 50 kHz. The diagnostic took observation for the ohmically heated plasma. The data was then processed to construct spatial and temporal profile of SXR intensity for Aditya plasma. This information was complimentary to the Silicon surface barrier detector (SBD) based array for the same plasma discharge. The cross calibration between the two was considerably satisfactory under the assumptions considered. (author)

  19. Development of real-time plasma analysis and control algorithms for the TCV tokamak using SIMULINK

    International Nuclear Information System (INIS)

    Felici, F.; Le, H.B.; Paley, J.I.; Duval, B.P.; Coda, S.; Moret, J.-M.; Bortolon, A.; Federspiel, L.; Goodman, T.P.; Hommen, G.; Karpushov, A.; Piras, F.; Pitzschke, A.; Romero, J.; Sevillano, G.; Sauter, O.; Vijvers, W.

    2014-01-01

    Highlights: • A new digital control system for the TCV tokamak has been commissioned. • The system is entirely programmable by SIMULINK, allowing rapid algorithm development. • Different control system nodes can run different algorithms at varying sampling times. • The previous control system functions have been emulated and improved. • New capabilities include MHD control, profile control, equilibrium reconstruction. - Abstract: One of the key features of the new digital plasma control system installed on the TCV tokamak is the possibility to rapidly design, test and deploy real-time algorithms. With this flexibility the new control system has been used for a large number of new experiments which exploit TCV's powerful actuators consisting of 16 individually controllable poloidal field coils and 7 real-time steerable electron cyclotron (EC) launchers. The system has been used for various applications, ranging from event-based real-time MHD control to real-time current diffusion simulations. These advances have propelled real-time control to one of the cornerstones of the TCV experimental program. Use of the SIMULINK graphical programming language to directly program the control system has greatly facilitated algorithm development and allowed a multitude of different algorithms to be deployed in a short time. This paper will give an overview of the developed algorithms and their application in physics experiments

  20. From profile to sawtooth control: developing feedback control using ECRH/ECCD systems on the TCV tokamak

    International Nuclear Information System (INIS)

    Paley, J I; Felici, F; Coda, S; Goodman, T P

    2009-01-01

    Real time control of heating systems is essential to maximize plasma performance and avoid or neutralize instabilities under changing plasma conditions. Several feedback control algorithms have been developed on the Tokamak a Configuration Variable (TCV) tokamak that use the electron cyclotron (ECRH/ECCD) system to control a wide range of plasma properties, including the plasma current, shape, profiles as well as the sawtooth instability. Controllers have been developed to obtain sawteeth of a pre-determined period, to maximize the sawtooth period using an extremum seeking control algorithm and finally to provide simultaneous control of the plasma emission profile peak and width using multiple independent EC actuators.

  1. Recent Progress on Spherical Torus Research

    Energy Technology Data Exchange (ETDEWEB)

    Ono, Masayuki [PPPL; Kaita, Robert [PPPL

    2014-01-01

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

  2. Development of Spherical Near Field Model for Geological Radioactive Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. Y.; Lee, K. J.; Chang, S. H. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, K. J.; Chang, S. H. [Khalifa Univ. of Science/Technology and Research, Abu Dhabi (United Arab Emirates)

    2012-03-15

    Modeling for geological radioactive waste repository can be divided into 3 parts. They are near field modeling related to engineered barrier, far field modeling related to natural barrier and biosphere modeling. In order to make the general application for safety assessment of geological waste repository, spherical geometry near field model has been developed. This model can be used quite extensively when users calculate equivalent spherical geometry for specific engineered barrier like equivalent waste radius, equivalent barrier radius and etc. Only diffusion was considered for general purpose but advection part can be updated. Goldsim and Goldsim Radionuclide Transport (RT) module were chosen and used as developing tool for the flexible modeling. Developer can freely make their own model with developer friendly graphic interface by using Goldsim. Furthermore, model with user friendly graphic interface can be developed by using Goldsim Dashboard Authoring module. The model has been validated by comparing the result with that of another model, inserting similar inputs and conditions. The model has been proved to be reasonably operating from the comparison result by validation process. Cylindrical model can be developed as a further work based on the knowledge and experience from this research.

  3. Development of Spherical Near Field Model for Geological Radioactive Waste Repository

    International Nuclear Information System (INIS)

    Kim, S. Y.; Lee, K. J.; Chang, S. H.; Lee, K. J.; Chang, S. H.

    2012-01-01

    Modeling for geological radioactive waste repository can be divided into 3 parts. They are near field modeling related to engineered barrier, far field modeling related to natural barrier and biosphere modeling. In order to make the general application for safety assessment of geological waste repository, spherical geometry near field model has been developed. This model can be used quite extensively when users calculate equivalent spherical geometry for specific engineered barrier like equivalent waste radius, equivalent barrier radius and etc. Only diffusion was considered for general purpose but advection part can be updated. Goldsim and Goldsim Radionuclide Transport (RT) module were chosen and used as developing tool for the flexible modeling. Developer can freely make their own model with developer friendly graphic interface by using Goldsim. Furthermore, model with user friendly graphic interface can be developed by using Goldsim Dashboard Authoring module. The model has been validated by comparing the result with that of another model, inserting similar inputs and conditions. The model has been proved to be reasonably operating from the comparison result by validation process. Cylindrical model can be developed as a further work based on the knowledge and experience from this research

  4. Research using small tokamaks

    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

  5. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

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

  6. Remote maintenance design activities and research and development accomplishments for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1988-01-01

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

  8. Development of a precise long-time digital integrator for magnetic measurements in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kurihara, Kenichi; Kawamata, Youichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-10-01

    Long-time D-T burning operation in a tokamak requires that a magnetic sensor must work in an environment of 14-MeV intense neutron field, and that the measurement system must output precise magnetic field values. A method of time-integration of voltage produced in a simple pick-up coil seems to have preferable features of good time response, easy maintenance, and resistance to neutron irradiation. However, an inevitably-produced signal drift makes it difficult to apply the method to the long-time integral operation. To solve this problem, we have developed a new digital integrator (a voltage-to-frequency converter and an up-down counter) with testing the trial boards in the JT-60 magnetic measurements. This reports all of the problems and their measures through the development steps in details, and shows how to apply this method to the ITER operation. (author)

  9. Development of SMM wave laser scattering apparatus for the measurements of waves and turbulences in the tokamak plasma

    International Nuclear Information System (INIS)

    Saito, T.; Hamada, Y.; Yamashita, T.; Ikeda, M.; Nakamura, M.

    1980-01-01

    The SMM wave laser scattering apparatus has been developed for the measurement of the waves and turbulences in the plasma. This apparatus will help greatly to clarify the physics of RF heating of the tokamak plasma. The present status of main parts of the apparatus, the SMM wave laser and the Schottky barrier diode mixer for the heterodyne receiver, are described. (author)

  10. Helical-type device and laser fusion. Rivals for tokamak-type device at n-fusion development in Japan

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Under the current policy on the research and development of nuclear fusion in Japan, as enunciated by the Atomic Energy Commission of Japan, the type of a prototype fusion reactor will be chosen after 2020 from tokamak, helical or some other type including the inertial confinement fusion using lasers. A prototype fusion reactor is the next step following the tokamak type International Thermonuclear Experimental Reactor (ITER). With the prototype reactor, the feasibility as a power plant will be examined. At present the main research and development of nuclear fusion in Japan are on tokamak type, which have been promoted by Japan Atomic Energy Research Institute (JAERI). As for the other types of nuclear fusion, researches have been carried out on the helical type in Kyoto University and National Institute for Fusion Science (NIFS), the mirror type in Tsukuba University, the tokamak type using superconductive coils in Kyushu University, and the laser fusion in Osaka University. The features and the present state of research and development of the Large Helical Device and the laser fusion which is one step away from the break-even condition are reported. (K.I.)

  11. The development of joining doped graphite to copper for first wall application in HT-7 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Zhangjian, E-mail: zhouzhj@mater.ustb.edu.cn [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Zhong Zhihong [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Chen Junling [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Ge Changchun [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China)

    2010-12-15

    Two joining methods have been developed for joining carbon based plasma facing material to copper based heat sink material for the potential application in HT-7 and EAST tokamak. The first joining method is based on brazing technique by using a rapidly solidified foil-type Ti-Zr based amorphous filler with a melting temperature of 850 deg. C. The other joining method is direct active metal casting-casting the premixed powders of copper and active transition metals on the mechanical machined carbon surface directly. SEM observations demonstrate high quality of joining surface for both joints. The brazing technique is more promising for fabrication joint with larger size compared with the direct active alloy casting method. High heat flux test using an e-beam device was performed on the actively cooled C/Cu joint fabricated by brazing method. There has no damage occurred on the joint after heat loading at 6 MW/m{sup 2}.

  12. Developing maintainability for tokamak fusion power systems. Phase II report. Volume I: executive summary

    International Nuclear Information System (INIS)

    Fuller, G.M.; Zahn, H.S.; Mantz, H.C.; Kaletta, G.R.; Waganer, L.M.; Carosella, L.A.; Conlee, J.L.

    1978-11-01

    The purpose of this report is to identify design features of fusion power reactors which contribute to the achievement of high levels of maintainability. Volume 1, the Executive Summary, presents the progress achieved toward this objective in this phase and includes a comparison with the results of the first phase study efforts. A series of maintainability design guidelines and an improved maintenance system are defined as initial steps in developing the requirements for a maintainable tokamak fusion power system. The principle comparative studies that are summarized include the determination of the benefits of various vacuum wall arrangements, the effect of unscheduled and scheduled maintenance of the first wall/blanket, some initial investigation of maintenance required for subsystems other than the first wall/blanket, and the impact of maintenance equipment failures

  13. Structure of magnetic field disturbances under development of disruptive instability in the ''Tokamak-6''

    International Nuclear Information System (INIS)

    Merezhkin, V.G.

    1978-01-01

    The structure and dynamics of disturbances of a poloidal field during the development of the breakaway instability in the Tokamak-6 are investigated. The behaviour of the symmetric and dipole field component, and the peculiarities of the structure of screw disturbances in a minor and major breakaways are analyzed. It was established that the structure of screw disturbances in minor breakaways is unchangeable and that the rearrangement in major breakaways is of a discrete nature. The relationship between the symmetric and screw components of disturbances of the poloidal field at the forward front of the disturbance increase was revealed. Data on the increments, scales and structure of screw disturbances, the ratios between the symmetric and screw components of field disturbances, and also on the magnitude of energy losses in typical breakaways are given

  14. Development of an alternating integrator for magnetic measurements for experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Liu, D. M., E-mail: dmliu@live.cn; Zhao, W. Z.; He, Y. G.; Chen, B. [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China); Wan, B. N.; Shen, B.; Huang, J.; Liu, H. Q. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2014-11-15

    A high-performance integrator is one of the key electronic devices for reliably controlling plasma in the experimental advanced superconducting tokamak for long pulse operation. We once designed an integrator system of real-time drift compensation, which has a low integration drift. However, it is not feasible for really continuous operations due to capacitive leakage error and nonlinearity error. To solve the above-mentioned problems, this paper presents a new alternating integrator. In the new integrator, the integrator system of real-time drift compensation is adopted as one integral cell while two such integral cells work alternately. To achieve the alternate function, a Field Programmable Gate Array built in the digitizer is utilized. The performance test shows that the developed integrator with the integration time constant of 20 ms has a low integration drift (<15 mV) for 1000 s.

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

    International Nuclear Information System (INIS)

    Gouge, M.J.; Combs, S.K.; Fisher, P.W.; Milora, S.L.

    1989-01-01

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

  16. A consistent formulation of wave propagation and conversion in low aspect ratio tokamaks with non-circular cross section

    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

  17. A study on the fusion reactor - Development of x-ray spectrometer for diagnosis of tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hong Young; Choi, Duk In; Seo, Sung Hun; Kwon, Gi Chung; Jun, Sang Jin; Heo, Sung Hoi; Lee, Chan Hui [Korea Advanced Institute of Science and Technolgoy, Taejon (Korea, Republic of)

    1996-09-01

    This report of research is on the development of X-ray Photo-Electron Spectrometer (PES) for diagnosis of tokamak plasma. The spectrometer utilizes the fact that the energy of photo-electron is given by the difference between the energy of X-ray and the binding energy of materials. In the research of this year, we constructed two spectrometers; one is operated in KAIST tokamak and the other in KT1 tokamak. In addition, we reviewed the characteristics of the x-ray filter, the photo-electric effect of carbon foils and the detection efficiency of MCP and x-ray radiation of plasma. We measured the x-ray radiation in tokamak and diagnosed the qualitative plasma parameters from the analysis of data. The major interesting plasma parameters, which we can diagnose with the spectrometer, are the electron temperature, Z{sub eff}, the spatial distribution of x-ray radiation and etc. 27 refs., 2 tabs., 20 figs. (author)

  18. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  19. The Development of Biomimetic Spherical Hydroxyapatite/Polyamide 66 Biocomposites as Bone Repair Materials

    Directory of Open Access Journals (Sweden)

    Xuesong Zhang

    2014-01-01

    Full Text Available A novel biomedical material composed of spherical hydroxyapatite (s-HA and polyamide 66 (PA biocomposite (s-HA/PA was prepared, and its composition, mechanical properties, and cytocompatibility were characterized and evaluated. The results showed that HA distributed uniformly in the s-HA/PA matrix. Strong molecule interactions and chemical bonds were presented between the s-HA and PA in the composites confirmed by IR and XRD. The composite had excellent compressive strength in the range between 95 and 132 MPa, close to that of natural bone. In vitro experiments showed the s-HA/PA composite could improve cell growth, proliferation, and differentiation. Therefore, the developed s-HA/PA composites in this study might be used for tissue engineering and bone repair.

  20. 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

  1. Tokamak physics

    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)

  2. 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

  3. Studies on the development of special graphite for use in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Bhatia, G.; Aggarwal, R.K.; Saha, M.; Sengupta, P.R.; Mishra, A. [National Physical Lab., New Delhi (India). Carbon Technology Unit

    2002-07-01

    Special graphite is considered as a critical component of the present-day tokamaks wherein it acts as the armour material for plasma-facing components. This graphite is required to possess, besides other characteristics, high values of bulk density, bending strength and electrical and thermal conductivities and a low value of ash content. Since such graphite was not commercially available in the country, efforts to develop it were initiated at the National Physical Laboratory, New Delhi. The basic approach to develop this graphite was based on green coke method of making the high density graphite, wherein the green coke was modified by incorporating in it small amounts of conducting carbon materials, i.e. needle coke, synthetic graphite and natural graphite. The resulting graphites were characterized with respect to various physical characteristics, namely, green density, weight loss, volume shrinkage, linear shrinkage, bulk density, bending strength, Young's modulus and electrical resistivity, etc. The results are described and discussed in the present paper. 6 refs., 2 tabs.

  4. Recent developments towards steady state physics and technology of tokamaks in Cadarache

    International Nuclear Information System (INIS)

    Jacquinot, J.G.

    2002-01-01

    Recently, Tore Supra has undergone a total change of internal components in order to upgrade the heat extraction capability to 25 MW for 1000 s, and address long pulse operation of a tokamak at a level of power density owing through the separatrix relevant for next step. The present paper will both give an overview of the experimental results obtained during the last campaigns and highlight the related technology developments: industrial realisation and tests with plasma of about 600 actively cooled plasma limiter components, new experimental results concerning heating and current drive systems (ECRH, ICRH, LHCD), injection of matter for long pulses (supersonic injection, high repetition rate pellet injection), stability and control of high confinement steady-state discharges sustained by the LH wave, theoretical and experimental investigations of electron heat transport. Highlights of technology developments directly applicable to ITER are also presented. Finally, a brief account is given of the European studies for validating Cadarache as a possible site for ITER, concluding that all ITER technical site requests are fully met. (author)

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

    International Nuclear Information System (INIS)

    Sebelin, E.

    1997-01-01

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

  6. An Analytic Approach to Developing Transport Threshold Models of Neoclassical Tearing Modes in Tokamaks

    International Nuclear Information System (INIS)

    Mikhailovskii, A.B.; Shirokov, M.S.; Konovalov, S.V.; Tsypin, V.S.

    2005-01-01

    Transport threshold models of neoclassical tearing modes in tokamaks are investigated analytically. An analysis is made of the competition between strong transverse heat transport, on the one hand, and longitudinal heat transport, longitudinal heat convection, longitudinal inertial transport, and rotational transport, on the other hand, which leads to the establishment of the perturbed temperature profile in magnetic islands. It is shown that, in all these cases, the temperature profile can be found analytically by using rigorous solutions to the heat conduction equation in the near and far regions of a chain of magnetic islands and then by matching these solutions. Analytic expressions for the temperature profile are used to calculate the contribution of the bootstrap current to the generalized Rutherford equation for the island width evolution with the aim of constructing particular transport threshold models of neoclassical tearing modes. Four transport threshold models, differing in the underlying competing mechanisms, are analyzed: collisional, convective, inertial, and rotational models. The collisional model constructed analytically is shown to coincide exactly with that calculated numerically; the reason is that the analytical temperature profile turns out to be the same as the numerical profile. The results obtained can be useful in developing the next generation of general threshold models. The first steps toward such models have already been made

  7. Development of FEMAG. Calculation code of magnetic field generated by ferritic plates in the tokamak devices

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Kazuhiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2003-03-01

    In design of the future fusion devises in which low activation ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect against the plasma as well as the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of magnetic field generated by ferritic steel. However iterative calculations concerning the non-linearity in B-H curve of ferritic steel disturbs high-speed calculation required as the design tool. In the strong toroidal magnetic field that is characteristic in the tokamak fusion devices, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward and any iteration calculation are unnecessary. Additionally objective ferritic steel geometry is limited to the thin plate and ferritic plates are installed along the toroidal magnetic field. Taking these special conditions into account, high-speed calculation code ''FEMAG'' has been developed. In this report, the formalization of 'FEMAG' code, how to use 'FEMAG', and the validity check of 'FEMAG' in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are numerical results of design studies for JT-60 modification. (author)

  8. The development of an in-vessel cryopump system for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Laughon, G.J.; Mahdavi, M.A.; Makariou, C.C.; Smith, J.P.; Schaffer, M.J.; Menon, M.M.

    1993-07-01

    The design, testing and initial operation of the DIII-D advanced divertor cryocondensation pumping system is presented. The pump resides inside the tokamak plasma containment vessel where it provides particle exhaust pumping, and it is subjected to Joule heating and hot particle heat loads during each 10 second discharge. In addition, the pump must withstand plasma disruption induced electromagnetic forces and 400 degrees C bake-out temperatures. Cooling is accomplished by forced flow liquid helium with the two-phase helium exhaust passing through a reliquefier for thermal efficiency. A prototype pump was constructed to study surface temperature rise as a function of flow geometry, applied heat load, helium mass flow rate, and pump outlet conditions. Prototype testing led to the development of a special geometry which was demonstrated to enhance two-phase flow stability and overall heat transfer. During initial operation, deuterium pumping speeds of 32,000 L/s at 2 mTorr pressure were achieved with a helium flow rate of 5 g/s. This speed was maintained during 300 W, 8 s long test beat pulses which meets operational goals

  9. Development of integrated insulation joint for cooling pipe in tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Abe, Tetsuya; Kawamura, Masashi; Yamazaki, Seiichiro.

    1994-08-01

    In a tokamak fusion reactor, an electrically insulated part is needed for an in-vessel piping system in order to break an electric circuit loop. When a closed loop is formed in the piping system, large induced electromagnetic forces during a plasma disruption (rapid plasma current quench) could give damages on the piping system. Ceramic brazing joint is a conventional method for the electric circuit break, but an application to the fusion reactor is not feasible due to its brittleness. Here, a stainless steel/ceramics/stainless steel functionally gradient material (FGM) has been proposed and developed as an integrated insulation joint of the piping system. Both sides of the joint can be welded to the main pipes, and expected to be reliable even in the fusion reactor environment. When the FGM joint is manufactured by way of a sintering process, a residual thermal stress is the key issue. Through detailed computations of the residual thermal stress and several trial productions, tubular elements of FGM joints have been successfully manufactured. (author)

  10. Development of burning plasma and advanced scenarios in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.

    2005-01-01

    Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q ∼ 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque. (author)

  11. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

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

  12. A better understanding of biomass co-firing by developing an advanced non-spherical particle tracking model

    DEFF Research Database (Denmark)

    Yin, Chungen; Rosendahl, Lasse Aistrup; Kær, Søren Knudsen

    2004-01-01

    -area-to-volume ratio and thus experiences a totally different motion and reaction as a non-spherical particle. Therefore, an advanced non-spherical particle-tracking model is developed to calculate the motion and reaction of nonspherical biomass particles. The biomass particles are assumed as solid or hollow cylinders......-gradient force. Since the drag and lift forces are both shape factor- and orientation-dependent, coupled particle rotation equations are resolved to update particle orientation. In the reaction of biomass particles, the actual particle surface area available and the average oxygen mass flux at particle surface...

  13. Development of wall conditioning and impurity monitoring systems in Versatile Experiment Spherical Torus (VEST)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.Y., E-mail: brbbebbero@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of); Yang, J.; Kim, Y.G.; Yang, S.M.; Kim, Y.S.; Lee, K.H. [Seoul National University, Seoul (Korea, Republic of); An, Y.H. [National Fusion Research Institute, Daejon (Korea, Republic of); Chung, K.J.; Na, Y.S. [Seoul National University, Seoul (Korea, Republic of); Hwang, Y.S., E-mail: yhwang@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of)

    2016-11-01

    Highlights: • The baking for partial wall heating and H{sub 2}/He GDC systems are developed in VEST. • The RGA and OES systems for monitoring impurities are constructed in VEST. • The partial baking and He GDC show limited effects on plasma characteristics. • H{sub 2} GDC above 4 h enables the longer plasma current duration up to ∼15 ms. • After H{sub 2} GDC, the discharge should be conducted within 3 h from treatment. - Abstract: Wall conditioning and impurity monitoring systems are developed in Versatile Experiment Spherical Torus (VEST). As a wall conditioning system, a baking system covering the vacuum vessel wall partially and a glow discharge cleaning (GDC) system using two electrodes with dc and 50 kHz power supplies are installed. The GDC system operates with hydrogen and helium gases for both chemical and physical desorption. The impurity monitoring system with residual gas analyzer (RGA), operating at <10{sup −5} Torr with a differential pumping system, is installed along with the optical emission spectroscopy (OES) system to monitor the hydrogen and impurity radiation lines. Effects of these wall conditioning techniques are investigated with the impurity monitoring system for ohmic discharges of VEST. The partial baking and He GDC show limited effects on plasma characteristics but sufficient H{sub 2} GDC above 4 h enables the longer plasma current duration up to ∼15 ms within 3 h from the end of treatment.

  14. Development of wall conditioning and impurity monitoring systems in Versatile Experiment Spherical Torus (VEST)

    International Nuclear Information System (INIS)

    Lee, H.Y.; Yang, J.; Kim, Y.G.; Yang, S.M.; Kim, Y.S.; Lee, K.H.; An, Y.H.; Chung, K.J.; Na, Y.S.; Hwang, Y.S.

    2016-01-01

    Highlights: • The baking for partial wall heating and H_2/He GDC systems are developed in VEST. • The RGA and OES systems for monitoring impurities are constructed in VEST. • The partial baking and He GDC show limited effects on plasma characteristics. • H_2 GDC above 4 h enables the longer plasma current duration up to ∼15 ms. • After H_2 GDC, the discharge should be conducted within 3 h from treatment. - Abstract: Wall conditioning and impurity monitoring systems are developed in Versatile Experiment Spherical Torus (VEST). As a wall conditioning system, a baking system covering the vacuum vessel wall partially and a glow discharge cleaning (GDC) system using two electrodes with dc and 50 kHz power supplies are installed. The GDC system operates with hydrogen and helium gases for both chemical and physical desorption. The impurity monitoring system with residual gas analyzer (RGA), operating at <10"−"5 Torr with a differential pumping system, is installed along with the optical emission spectroscopy (OES) system to monitor the hydrogen and impurity radiation lines. Effects of these wall conditioning techniques are investigated with the impurity monitoring system for ohmic discharges of VEST. The partial baking and He GDC show limited effects on plasma characteristics but sufficient H_2 GDC above 4 h enables the longer plasma current duration up to ∼15 ms within 3 h from the end of treatment.

  15. Gas Electron Multipliers: Development of large area GEMs and spherical GEMs

    CERN Document Server

    Duarte Pinto, Serge; Brock, Ian

    2011-01-01

    Gaseous radiation detectors have been a crucial part of high-energy physics instrumentation since the 1960s, when the first multiwire proportional counters were built. In the 1990s the first micropattern gas detectors (MPGDs) saw the light; with sub-millimeter feature sizes these novel detectors were faster and more accurate than their predecessors. The gas electron multiplier (GEM) is one of the most successful of these technologies. It is a charge multiplication structure made from a copper clad polymer foil, pierced with a regular and dense pattern of holes. I will describe the properties and the application of GEMs and GEM detectors, and the research and development I have done on this technology. Two of the main objectives were the development of large area GEMs (~m^2) for particle physics experiments and GEMs with a spherical shape for x-ray or neutron diffraction detectors. Both have been realized, and the new techniques involved are finding their way to applications in research and industry.

  16. Gas electron multipliers. Development of large area GEMS and spherical GEMS

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Serge Duarte

    2011-08-15

    Gaseous radiation detectors have been a crucial part of high-energy physics instrumentation since the 1960s, when the first multiwire proportional counters were built. In the 1990s the first micropattern gas detectors (MPGDS) saw the light; with sub-millimeter feature sizes these novel detectors were faster and more accurate than their predecessors. The gas electron multiplier (GEM) is one of the most successful of these technologies. It is a charge multiplication structure made from a copper clad polymer foil, pierced with a regular and dense pattern of holes. I describe the properties and the application of GEMs and GEM. detectors, and the research and development I have done on this technology. Two of the main objectives were the development of large area GEMs ({proportional_to}m{sup 2}) for particle physics experiments and GEMs with a spherical shape for X-ray or neutron diffraction detectors. Both have been realized, and the new techniques involved are finding their way to applications in research and industry. (orig.)

  17. Gas electron multipliers: Development of large area GEMS and spherical GEMS

    International Nuclear Information System (INIS)

    Pinto, Serge Duarte

    2011-08-01

    Gaseous radiation detectors have been a crucial part of high-energy physics instrumentation since the 1960s, when the first multiwire proportional counters were built. In the 1990s the first micropattern gas detectors (MPGDS) saw the light; with sub-millimeter feature sizes these novel detectors were faster and more accurate than their predecessors. The gas electron multiplier (GEM) is one of the most successful of these technologies. It is a charge multiplication structure made from a copper clad polymer foil, pierced with a regular and dense pattern of holes. I describe the properties and the application of GEMs and GEM. detectors, and the research and development I have done on this technology. Two of the main objectives were the development of large area GEMs (∝m 2 ) for particle physics experiments and GEMs with a spherical shape for X-ray or neutron diffraction detectors. Both have been realized, and the new techniques involved are finding their way to applications in research and industry. (orig.)

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

    Czech Academy of Sciences Publication Activity Database

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

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

  19. Study of a spherical torus based volumetric neutron source for nuclear technology testing and development

    International Nuclear Information System (INIS)

    Cheng, E.T.; Cerbone, R.J.; Sviatoslavsky, I.N.; Galambos, L.D.; Peng, Y.-K.M.

    2000-01-01

    A plasma based, deuterium and tritium (DT) fueled, volumetric 14 MeV neutron source (VNS) has been considered as a possible facility to support the development of the demonstration fusion power reactor (DEMO). It can be used to test and develop necessary fusion blanket and divertor components and provide sufficient database, particularly on the reliability of nuclear components necessary for DEMO. The VNS device can be complement to ITER by reducing the cost and risk in the development of DEMO. A low cost, scientifically attractive, and technologically feasible volumetric neutron source based on the spherical torus (ST) concept has been conceived. The ST-VNS, which has a major radius of 1.07 m, aspect ratio 1.4, and plasma elongation three, can produce a neutron wall loading from 0.5 to 5 MW m -2 at the outboard test section with a modest fusion power level from 38 to 380 MW. It can be used to test necessary nuclear technologies for fusion power reactor and develop fusion core components include divertor, first wall, and power blanket. Using staged operation leading to high neutron wall loading and optimistic availability, a neutron fluence of more than 30 MW year m -2 is obtainable within 20 years of operation. This will permit the assessments of lifetime and reliability of promising fusion core components in a reactor relevant environment. A full scale demonstration of power reactor fusion core components is also made possible because of the high neutron wall loading capability. Tritium breeding in such a full scale demonstration can be very useful to ensure the self-sufficiency of fuel cycle for a candidate power blanket concept

  20. Development and experimental evaluation of theoretical models for ion cyclotron resonance frequency heating of tokamak plasmas

    International Nuclear Information System (INIS)

    Mantsinen, M.

    1999-01-01

    Heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is a well-established method for auxiliary heating of present-day tokamak plasmas and is envisaged as one of the main heating techniques for the International Thermonuclear Experimental Reactor (ITER) and future reactor plasmas. In order to predict the performance of ICRF heating in future machines, it is important to benchmark present theoretical modelling with experimental results on present tokamaks. This thesis reports on development and experimental evaluation of theoretical models for ICRF heating at the Joint European Torus (JET). Several ICRF physics effects and scenarios have been studied. Direct importance to the ITER is the theoretical analysis of ICRF heating experiments with deuterium-tritium (D-T) plasmas. These experiments clearly demonstrate the potential of ICRF heating for auxiliary heating of reactor plasmas. In particular, scenarios with potential for good bulk ion heating and enhanced D-T fusion reactivity have been identified. Good bulk ion heating is essential for reactor plasmas in order to obtain a high ion temperature and a high fusion reactivity. In JET good bulk ion heating with ICRF waves has been achieved in high-performance discharges by adding ICRF heating to neutral beam injection. In these experiments, as in other JET discharges where damping at higher harmonics of the ion cyclotron frequency takes place, so-called finite Larmor radius (FLR) effects play an important role. Due to FLR effects, the resonating ion velocity distribution function can have a strong influence on the power deposition. Evidence for this effect has been obtained from the third harmonic deuterium heating experiments. Because of FLR effects, the wave-particle interaction can also become weak at certain ion energies, which prevents resonating ions from reaching higher energies. When interacting with the wave, an ion receives not only a change in energy but also a change in

  1. Development and experimental evaluation of theoretical models for ion cyclotron resonance frequency heating of tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Mantsinen, M. [Helsinki Univ. of Technology, Espoo (Finland). Dept. of Technical Physics

    1999-06-01

    Heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is a well-established method for auxiliary heating of present-day tokamak plasmas and is envisaged as one of the main heating techniques for the International Thermonuclear Experimental Reactor (ITER) and future reactor plasmas. In order to predict the performance of ICRF heating in future machines, it is important to benchmark present theoretical modelling with experimental results on present tokamaks. This thesis reports on development and experimental evaluation of theoretical models for ICRF heating at the Joint European Torus (JET). Several ICRF physics effects and scenarios have been studied. Direct importance to the ITER is the theoretical analysis of ICRF heating experiments with deuterium-tritium (D-T) plasmas. These experiments clearly demonstrate the potential of ICRF heating for auxiliary heating of reactor plasmas. In particular, scenarios with potential for good bulk ion heating and enhanced D-T fusion reactivity have been identified. Good bulk ion heating is essential for reactor plasmas in order to obtain a high ion temperature and a high fusion reactivity. In JET good bulk ion heating with ICRF waves has been achieved in high-performance discharges by adding ICRF heating to neutral beam injection. In these experiments, as in other JET discharges where damping at higher harmonics of the ion cyclotron frequency takes place, so-called finite Larmor radius (FLR) effects play an important role. Due to FLR effects, the resonating ion velocity distribution function can have a strong influence on the power deposition. Evidence for this effect has been obtained from the third harmonic deuterium heating experiments. Because of FLR effects, the wave-particle interaction can also become weak at certain ion energies, which prevents resonating ions from reaching higher energies. When interacting with the wave, an ion receives not only a change in energy but also a change in

  2. 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)

  3. Forthcoming Break-Even Conditions of Tokamak Plasma Performance for Fusion Energy Development

    Science.gov (United States)

    Hiwatari, Ryoji; Okano, Kunihiko; Asaoka, Yoshiyuki; Tokimatsu, Koji; Konishi, Satoshi; Ogawa, Yuichi

    The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value ΒN), confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fnGW), the electric break-even condition requires the simultaneous achievement of 1.2 market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of Btmax = 16 T, ηe = 40 %, plant availability 60 %, and a radial build with/without CS coil, the economic break-even condition requires ΒN ˜ 5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with ΒN ˜ 3.0 in the ITER project leads to the upper region of the break-even price in the present world energy scenario, which implies that it is necessary to improve the plasma performance beyond that of the ITER advanced plasma operation.

  4. 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

  5. Review of ICRF antenna development and heating experiments up to advanced experiment I, 1989 on the JT-60 tokamak

    International Nuclear Information System (INIS)

    Fujii, Tsuneyuki

    1992-03-01

    Two main subjects of ion cyclotron range of frequencies (ICRF) heating on JT-60 are described in this paper from development phase of the JT-60 ICRF heating system up to advanced experiment I, 1989. One is antenna design and development for the high power JT-60 ICRF heating system (6 MW for 10 s at a frequency range of 108 - 132 MHz). The other is the experimental investigation of characteristics of second harmonic ICRF heating in a large tokamak. (J.P.N.)

  6. Plasma Fluctuation Studies in the TCV Tokamak: Modeling of Shaping Effects and Advanced Diagnostic Development

    International Nuclear Information System (INIS)

    Marinoni, A.

    2009-10-01

    One of the most important issues for magnetic-confinement fusion research is the so-called anomalous transport across magnetic field lines, i.e. transport that is in excess of that caused by collisional processes. The need to reduce anomalous transport in order to increase the efficiency of a prospective fusion reactor must be addressed through an investigation of its fundamental underlying causes. This thesis is divided into two distinct components: one experimental and instrumental, and the other theoretical and based on numerical modeling. The experimental part consists of the design and installation of a new diagnostic for core turbulence fluctuations in the TCV tokamak. An extensive conceptual investigation of a number of possible solutions, including Beam Emission Spectroscopy, Reflectometry, Cross Polarization, Collective Scattering and different Imaging techniques, was carried out at first. A number of criteria, such as difficulties in data interpretation, costs, variety of physics issues that could be addressed and expected performance, were used to compare the different techniques for specific application to the TCV tokamak. The expected signal to noise ratio and the required sampling frequency for TCV were estimated on the basis of a large number of linear, local gyrokinetic simulations of plasma fluctuations. This work led to the choice of a Zernike phase contrast imaging system in a tangential launching configuration. The diagnostic was specifically designed to provide information on turbulence features up to now unknown. In particular, it is characterized by an outstanding spatial resolution and by the capability to measure a very broad range of fluctuations, from ion to electron Larmor radius scales, thus covering the major part of the instabilities expected to be at play in TCV. The spectrum accessible covers the wavenumber region from 0.9 cm -1 to 60 cm -1 at 24 radial positions with 3 MHz bandwidth. The diagnostic is an imaging technique and is

  7. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

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

  8. 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

  9. 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

  10. Spherical CNNs

    OpenAIRE

    Cohen, Taco S.; Geiger, Mario; Koehler, Jonas; Welling, Max

    2018-01-01

    Convolutional Neural Networks (CNNs) have become the method of choice for learning problems involving 2D planar images. However, a number of problems of recent interest have created a demand for models that can analyze spherical images. Examples include omnidirectional vision for drones, robots, and autonomous cars, molecular regression problems, and global weather and climate modelling. A naive application of convolutional networks to a planar projection of the spherical signal is destined t...

  11. 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

  12. 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

  13. ARIES tokamak reactor study

    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

  14. 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

  15. 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

  16. Forthcoming break-even conditions of tokamak plasma performance for fusion energy development

    International Nuclear Information System (INIS)

    Hiwatari, Ryoji; Okano, Kunihiko; Asaoka, Yoshiyuki; Tokimatsu, Koji; Konishi, Satoshi; Ogawa, Yuichi

    2005-01-01

    The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value β N , confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fn GW ), the electric break-even condition requires the simultaneous achievement of 1.2 N GW tmax =16 T, thermal efficiency η e =30%, and current drive power P NBI N ∼1.8, HH≠1.0, and fn GW ∼0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric break-even condition. The second condition is the economic break-even condition, which is required for fusion energy to be selected as an alternative energy source in the energy market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO 2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of B tmax =16 T, η e =40%, plant availability 60%, and a radial build with/without CS coil, the economic break-even condition requires β N ∼5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with β N ∼3.0 in the ITER project leads to the upper region of the break

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sebelin, E

    1997-12-15

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

  19. Development and application of poloidal correlation reflectometry to study turbulent structures in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Prisiazhniuk, Dmitrii

    2017-06-05

    One of the key question of high temperature plasma confinement in a magnetic field is how plasma turbulence influences the radial transport of particles and energy. A better understanding of transport processes caused by turbulence would allow to improve the plasma confinement in fusion devices. To this end a deeper understanding of the mechanisms controlling the development, saturation and stabilization of turbulence is needed. From the experimental point of view a main challenge in these investigations is the measurement of plasma parameters on both small temporal (μs) and spatial (mm) scales. In this thesis a new microwave heterodyne poloidal correlation reflectometry diagnostic has been developed and installed at the ASDEX Upgrade tokamak to investigate the cross-correlation of turbulent density fluctuations. This diagnostic yields information on fundamental turbulence parameters such as the perpendicular propagation velocity v {sub perpendicular} {sub to}, the perpendicular correlation length l {sub perpendicular} {sub to} (characteristic size of the turbulent eddies) and the decorrelation time τ{sub d} (characteristic life time of the turbulent eddies) over a wide range of plasma densities. The inclination of the turbulent eddies α in the poloidal-toroidal plane spanned by the magnetic flux surfaces of a tokamak, being a measure of the magnetic field pitch angle, can also be obtained. The turbulence investigations were performed in low confinement mode (L-mode) plasmas for a range of plasma parameters. All measurements were interpreted taking into account the transfer function of reflectometry in the Born approximation. The results are compared with theoretical predictions and simulations. In the first part of this thesis the inclination and the propagation of turbulent structures are investigated. It is shown that eddies are nearly aligned to the magnetic field line and, therefore, the magnetic field pitch angle can be measured with a precision of about 1

  20. Development of plasma diagnostics technologies - Measurement of transport= parameters in tokamak edge plasma by using electric transport probes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kyu Sun; Chang, Do Hee; Sim, Yeon Gun; Kim, Jin Hee [Hanyang University, Seoul (Korea, Republic of)

    1995-08-01

    Electric transport probe system is developed for the measurement of electron temperature, floating potential, plasma density and flow velocity of= edge plasmas in the KT-2 medium size tokamak. Experiments have been performed in KT-1 small size tokamak. Electric transport probe is composed of a single probe(SP) and a Mach probe (MP). SP is used for the measurements of electron density, floating potential, and plasma density and measured values are {approx} 3*10{sup 11}/cm{sup -3}, -20 volts, 15 {approx} 25 eV. For the most discharges, respectively. MP is for the measurements of toroidal(M{sub T}) and poloidal(M{sub P}) flow velocities, and density, which are M{sub T} {approx_equal} .0.85, M{sub P} {approx_equal}. 0.17, n. {approx_equal} 2.1*10{sup 11} cm{sup -3}, respectively. A triple probe is also developed for the direct reading of T{sub e} and n{sub e}, and is used for DC, RF, and RF+DC plasma in APL of Hanyang university. 38 refs., 36 figs. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-01

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

  2. Tokamak Simulation Code modeling of NSTX

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  3. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

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

    International Nuclear Information System (INIS)

    Lee, V.D.

    1987-01-01

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

  5. Development of plasma current waveform adjusting system ZLJ for tokamak device HL-1

    International Nuclear Information System (INIS)

    Wang Shangbing; Hu Haotian; Tang Fangqun; Zhou Yongzheng; Chu Xiuzhong; Cheng Jiashun; Gao Yunxia

    1989-12-01

    The control of some typical Tokamak discharge waveforms has been achieved by using plasma current waveform adjusting system ZLJ in the ohmic heating of HL-1. The discharge waveforms include a series of regular plasma current waveforms with various slow rising rate, such as 80 kA, 450 ms long flat-topping; 100 kA, 200 ms rising; 200 ms falt-topping and 180 kA, 400 ms slow rising etc. The design principle of the system and the initial experimental results are described

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

    Science.gov (United States)

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

    2018-03-01

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

  7. Development of high-speed and wide-angle visible observation diagnostics on Experimental Advanced Superconducting Tokamak using catadioptric optics

    International Nuclear Information System (INIS)

    Yang, J. H.; Hu, L. Q.; Zang, Q.; Han, X. F.; Shao, C. Q.; Sun, T. F.; Chen, H.; Wang, T. F.; Li, F. J.; Hu, A. L.; Yang, X. F.

    2013-01-01

    A new wide-angle endoscope for visible light observation on the Experimental Advanced Superconducting Tokamak (EAST) has been recently developed. The head section of the optical system is based on a mirror reflection design that is similar to the International Thermonuclear Experimental Reactor-like wide-angle observation diagnostic on the Joint European Torus. However, the optical system design has been simplified and improved. As a result, the global transmittance of the system is as high as 79.6% in the wavelength range from 380 to 780 nm, and the spatial resolution is <5 mm for the full depth of field (4000 mm). The optical system also has a large relative aperture (1:2.4) and can be applied in high-speed camera diagnostics. As an important diagnostic tool, the optical system has been installed on the HT-7 (Hefei Tokamak-7) for its final experimental campaign, and the experiments confirmed that it can be applied to the investigation of transient processes in plasma, such as ELMy eruptions in H-mode, on EAST

  8. The Tokamak IST-TOK

    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

  9. Design, Development and Testing of Inconel Alloy IN718 Spherical Gas Bottle for Oxygen Storage

    Science.gov (United States)

    Chenna Krishna, S.; Agilan, M.; Sudarshan Rao, G.; Singh, Satish Kumar; Narayana Murty, S. V. S.; Venkata Narayana, Ganji; Beena, A. P.; Rajesh, L.; Jha, Abhay K.; Pant, Bhanu

    2017-11-01

    This paper describes the details of design, manufacture and testing of 200 mm diameter spherical gas bottle of Inconel 718 (IN718) with nominal wall thickness of 2.3 mm. Gas bottle was designed for the specified internal pressure loading with a thickness of 2.9 mm at the circumferential weld which was brought down to 2.3 mm at the membrane locations. Hemispherical forgings produced through closed-die hammer forging were machined and electron beam welded to produce a spherical gas bottle. Duly welded gas bottle was subjected to standard aging treatment to achieve the required tensile strength. Aged gas bottle was inspected for dimensions and other stringent quality requirements using various nondestructive testing techniques. After inspection, gas bottle was subjected to pressure test for maximum expected operating pressure and proof pressure of 25 and 37.5 MPa, respectively. Strain gauges were bonded at different locations on the gas bottle to monitor the strains during the pressure test and correlated with the predicted values. The predicted strain matched well with the experimental strain confirming the design and structural integrity.

  10. Development of the α-decay theory of spherical nuclei by means of the shell model

    International Nuclear Information System (INIS)

    Holan, S.

    1978-01-01

    The new results achieved within the α-decay theory of spherical nuclei with a (2)-(5) integral formula, unaffected by arbitrary parameters, taking into account the finite shape of the α particle and using a basis of Woods-Saxon uniparticle functions to describe initial and final nuclei, may be summarized as follows: Through α-width calculations performed for many spherical nuclei it has been proved that experimental classifying of α-transition into favoured and unfavoured transitions as well as the hyperfine structure of the transitions can be theoretically explained if considered the nucleon-nucleon correlations in the description of initial and final nuclei; The absolute values of the theoretical α-widths obtained are about 10 2 times smaller compared to the experimental ones. This might be due to an oversimplified approximation of the α-particle-daughter nucleus interaction potential or either to an inaccuracy of the model functions used in describing nucleus decay in the surface area. (author)

  11. Recent developments on the 110 GHz electron cyclotron installation on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Ponce, D.; Callis, R.W.; Cary, W.P.; Ferron, J.R.; Green, M.; Grunloh, H.J.; Gorelov, Y.; Lohr, J.; Ellis, R.A.

    2003-01-01

    Significant improvements are being implemented to the capability of the 110 GHz electron cyclotron system on the DIII-D tokamak. Chief among these is the addition of the fifth and sixth 1 MW class gyrotrons, increasing the power available for auxiliary heating and current drive by nearly 60%. These tubes use artificially grown diamond r.f. output windows to obtain high power with long pulse capability. The beams from these tubes are nearly Gaussian, facilitating coupling to the waveguide. A new fully articulating dual launcher capable of high speed spatial scanning has been designed and tested. The launcher has two axis independent steering for each waveguide. The mirrors can be rotated at up to 100 deg./s. A new feedback system linking the DIII-D Plasma Control System (PCS) with the gyrotron beam voltage waveform generators permits real-time feedback control of some plasma properties such as electron temperature. The PCS can use a variety of plasma monitors to generate its control signal, including electron cyclotron emission and Mirnov probes. Electron cyclotron heating and electron cyclotron current drive were used during this year's DIII-D experimental campaign to control electron temperature, density, and q profiles, induce an ELM-free H-mode, and suppress the m=2/n=1 neoclassical tearing mode. The new capabilities have expanded the role of EC systems in tokamak plasma control

  12. Progress Towards High Performance, Steady-state Spherical Torus

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.; Bigelow, T.; Bitter, M.; Blanchard, W.; Boedo, J.; Bourdelle, C.; Bush, C.; Choe, W.; Chrzanowski, J.; Darrow, D.S.; Diem, S.J.; Doerner, R.; Efthimion, P.C.; Ferron, J.R.; Fonck, R.J.; Fredrickson, E.D.; Garstka, G.D.; Gates, D.A.; Gray, T.; Grisham, L.R.; Heidbrink, W.; Hill, K.W.; Hoffman, D.; Jarboe, T.R.; Johnson, D.W.; Kaita, R.; Kaye, S.M.; Kessel, C.; Kim, J.H.; Kissick, M.W.; Kubota, S.; Kugel, H.W.; LeBlanc, B.P.; Lee, K.; Lee, S.G.; Lewicki, B.T.; Luckhardt, S.; Maingi, R.; Majeski, R.; Manickam, J.; Maqueda, R.; Mau, T.K.; Mazzucato, E.; Medley, S.S.; Menard, J.; Mueller, D.; Nelson, B.A.; Neumeyer, C.; Nishino, N.; Ostrander, C.N.; Pacella, D.; Paoletti, F.; Park, H.K.; Park, W.; Paul, S.F.; Peng, Y.-K. M.; Phillips, C.K.; Pinsker, R.; Probert, P.H.; Ramakrishnan, S.; Raman, R.; Redi, M.; Roquemore, A.L.; Rosenberg, A.; Ryan, P.M.; Sabbagh, S.A.; Schaffer, M.; Schooff, R.J.; Seraydarian, R.; Skinner, C.H.; Sontag, A.C.; Soukhanovskii, V.; Spaleta, J.; Stevenson, T.; Stutman, D.; Swain, D.W.; Synakowski, E.; Takase, Y.; Tang, X.; Taylor, G.; Timberlake, J.; Tritz, K.L.; Unterberg, E.A.; Von Halle, A.; Wilgen, J.; Williams, M.; Wilson, J.R.; Xu, X.; Zweben, S.J.; Akers, R.; Barry, R.E.; Beiersdorfer, P.; Bialek, J.M.; Blagojevic, B.; Bonoli, P.T.; Carter, M.D.; Davis, W.; Deng, B.; Dudek, L.; Egedal, J.; Ellis, R.; Finkenthal, M.; Foley, J.; Fredd, E.; Glasser, A.; Gibney, T.; Gilmore, M.; Goldston, R.J.; Hatcher, R.E.; Hawryluk, R.J.; Houlberg, W.; Harvey, R.; Jardin, S.C.; Hosea, J.C.; Ji, H.; Kalish, M.; Lowrance, J.; Lao, L.L.; Levinton, F.M.; Luhmann, N.C.; Marsala, R.; Mastravito, D.; Menon, M.M.; Mitarai, O.; Nagata, M.; Oliaro, G.; Parsells, R.; Peebles, T.; Peneflor, B.; Piglowski, D.; Porter, G.D.; Ram, A.K.; Rensink, M.; Rewoldt, G.; Roney, P.; Shaing, K.; Shiraiwa, S.; Sichta, P.; Stotler, D.; Stratton, B.C.; Vero, R.; Wampler, W.R.; Wurden, G.A.

    2003-01-01

    Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction (∼60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted

  13. Development of ion diagnostic system based on electrostatic probe in the boundary plasma of the JFT-2M tokamak

    International Nuclear Information System (INIS)

    Uehara, Kazuya; Kawakami, Tomohide; Amemiya, Hiroshi; Hoethker, K.; Cosler, A.; Bieger, W.

    1995-06-01

    An ion diagnostic system using electrostatic probes for measurements in the JFT-2M tokamak boundary plasma has been developed under the collaboration program between KFA and JAERI. The rotating double probe system, on which the Hoethker double probe and Amemiya asymmetric probe can mounted, are manufactured at KFA workshop while the linear driver to support the rotating double probe, the ion toothbrush probe, the Katsumata probe and the cubic Mach probe are developed at JAERI. This report describes the hardware of this probe system for ion diagnostics in the boundary plasma and preliminary data obtained by means of this system. Furthermore, results on the transport are estimated on the basis of these probe data. (author)

  14. Development of Tokamak reactor system code and conceptual studies of DEMO with He Cooled Molten Li blanket

    International Nuclear Information System (INIS)

    Hong, B.G.; Lee, Dong Won; Kim, Yong Hi

    2007-01-01

    To develop the concepts of fusion power plants and identify the design parameters, we have been developing the tokamak reactor system code. The system code can take into account a wide range of plasma physics and technology effects simultaneously and it can be used to find design parameters which optimize the given figure of merits. The outcome of the system studies using the system code is to identify which areas of plasma physics and technologies and to what extent should be developed for realization of a given fusion power plant concepts. As an application of the tokamak reactor system code, we investigate the performance of DEMO for early realization with a limited extension from the plasma physics and technology used in the design of the ITER. Main requirements for DEMO are selected as: 1) to demonstrate tritium self-sufficiency, 2) to generate net electricity, and 3) for steady-state operation. The size of plasma is assumed to be same as that of ITER and the plasma parameters which characterize the performance, i.e. normalized β value, β N , confinement improvement factor for the H-mode, H and the ratio of the Greenwald density limit n/n G are assumed to be improved beyond those of ITER: β N >2.0, H>1.0 and n/n G >1.0. Tritium self-sufficiency is provided by the He Cooled Molten Lithium (HCML) blanket with the total thickness of 2.5 m including the shield. With n/n G >1.2, net electric power bigger than 500 MW is possible with β N >4.0 andH>1.2. To access operation space for higher electric power, main restrictions are given by the divertor heat load and the steady-state operation requirements. Developments in both plasma physics and technology are required to handle high heat load and to increase the current drive efficiency. (orig.)

  15. 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)

  16. New development of JFT-2M Tokamak (3) data processing system

    International Nuclear Information System (INIS)

    Fukuchi, Y.; Oyabu, I.; Hirose, T.; Ichimura, H.; Inoue, K.; Komoto, Y.

    1986-01-01

    A data acquisition system for JFT-2M Tokamak is a computer complex system consisting of a CAMAC serial highway, a front-end computer, and a main computer, which are ranked in a definite hierarchical structure. This paper reports the data processing system by the main computer (using a super-mini-computer MELCOM 70/250) which is situated on the highest level in the data acquisition system and performs unified management and control over the system. The features of the data processing system by the main computer are as follows: (1) Expandability of the system based on the definite hierarchical structure; (2) Five-dimensional multi-processing (setup, acquisition, analysis, display, and storage); (3) Realization of RAS (Reliability, Availability, and Serviceability) function; and (4) Easy-to-use man-machine interface that provides: flexibility in CAMAC system configuration, open-ended interface and file history managing

  17. Developments in the theory of trapped particle pressure gradient driven turbulence in tokamaks and stellarators

    International Nuclear Information System (INIS)

    Diamond, P.H.; Biglari, H.; Gang, F.Y.

    1991-01-01

    Recent advances in the theory of trapped particle pressure gradient driven turbulence are summarized. A novel theory of trapped ion convective cell turbulence is presented. It is shown that non-linear transfer to small scales occurs, and that saturation levels are not unphysically large, as previously thought. As the virulent saturation mechanism of ion Compton scattering is shown to result in weak turbulence at higher frequencies, it is thus likely that trapped ion convective cells are the major agent of tokamak transport. Fluid like trapped electron modes at short wavelengths (k θ ρ i > 1) are shown to drive an inward particle pinch. The characteristics of convective cell turbulence in flat density discharges are described, as is the stability of dissipative trapped electron modes in stellarators, with flexible magnetic field structure. The role of cross-correlations in the dynamics of multifield models of drift wave turbulence is discussed. (author). 32 refs, 8 figs, 1 tab

  18. Recent developments in the role of atomic processes in future tokamaks

    International Nuclear Information System (INIS)

    Post, D.

    1996-01-01

    Since the beginning of magnetic fusion research, reducing the impurity level in experiments has been strongly correlated with successful achievement of high performance plasmas. One of the most important examples of this was the recognition that the use of tungsten as a plasma facing material and the associated high radiative losses were responsible for the poor performance of the ORMAK and PLT tokamaks. Tungsten was replaced with graphite and the central plasma temperature in PLT increased a factor of ten. The magnetic fusion program is now planning on constructing an ignited fusion experiment. One of the major design issues is the reduction of the peak heat loads on the plasma facing components. It appears that the carefully controlled introduction of impurities can lead to a solution of the problem. copyright 1996 American Institute of Physics

  19. Development of a positronium atom diagnostic beam to study transport in tokamaks

    International Nuclear Information System (INIS)

    Surko, C.M.

    1992-01-01

    Anomalous transport is probably the key physics issue in magnetic confinement fusion. It severely limits our ability to create and maintain a dense, hot, fusion plasma. There is also good evidence that there is a fundamental lack of understanding in this area, namely the transport of particles and energy induced by waves and fluctuations in magnetically confined plasmas. The positronium atom beam diagnostic, if successful, can provide a qualitatively new and different way of studying transport in tokamak fusion plasmas. The use of the positron as a thermalized, electron-mass test particle will allow important new tests of current theories of plasma transport. In particular, it could provide unique insights into the potential role of magnetic fluctuations in producing anomalous transport. This is particularly significant in that, at present, there is essentially no in situ probe of these fluctuations or the transport which they produce. Some results of this study are summarized

  20. Development and performance of high speed processing system of magnetohydrodynamic equilibria for discharge analyses on the J T-60 tokamak

    International Nuclear Information System (INIS)

    Hasegawa, Yukihiro; Nakamura, Yukiharu; Shirai, Hiroshi; Hamamatsu, Kiyotaka; Harada, Yoshio; Kikuchi, Mitsuru; Nakata, Yoshihiro

    1999-01-01

    In order to provide a set of magnetohydrodynamic (MHD) equilibrium database which is indispensable for both the studies on improvement of energy confinement and stabilization of MHD activities in tokamaks, a high speed data-processing system synchronizing with J T-60 discharge sequence was newly developed by utilizing the latest model of hugh speed workstation and by optimizing the parallel processing technique to perform fast calculation of MHD equilibria. This high speed system was found to have a sufficient ability to complete the whole equilibrium calculations during each inter-shot period. Cooperating with the mass data storage subsystem preserving the latest equilibrium database automatically, the animated discharge monitoring subsystem provides valuable information for the J T-60 operator to determine control parameters of the succeeding discharge. This report describes the system performance realized in the J T-60 experiment. (author)

  1. Development of laser-based technology for the routine first wall diagnostic on the tokamak EAST: LIBS and LIAS

    Science.gov (United States)

    Hu, Z.; Gierse, N.; Li, C.; Liu, P.; Zhao, D.; Sun, L.; Oelmann, J.; Nicolai, D.; Wu, D.; Wu, J.; Mao, H.; Ding, F.; Brezinsek, S.; Liang, Y.; Ding, H.; Luo, G.; Linsmeier, C.; EAST Team

    2017-12-01

    A laser based method combined with spectroscopy, such as laser-induced breakdown spectroscopy (LIBS) and laser-induced ablation spectroscopy (LIAS), is a promising technology for plasma-wall interaction studies. In this work, we report the development of in situ laser-based diagnostics (LIBS and LIAS) for the assessment of static and dynamic fuel retention on the first wall without removing the tiles between and during plasma discharges in the Experimental Advanced Superconducting Tokamak (EAST). The fuel retention on the first wall was measured after different wall conditioning methods and daily plasma discharges by in situ LIBS. The result indicates that the LIBS can be a useful tool to predict the wall condition in EAST. With the successful commissioning of a refined timing system for LIAS, an in situ approach to investigate fuel retention is proposed.

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

    International Nuclear Information System (INIS)

    1979-02-01

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

  3. Development of a single cell spherical shell model for an investigation of electrical properties with a computing program

    Directory of Open Access Journals (Sweden)

    Boonlamp, M.

    2005-03-01

    Full Text Available A spherical double shell model (SDM for a single cell has been developed, using Laplace’s equation in spherical coordinates and boundary conditions. Electric field intensities and dielectric constants of each region inside and outside of the cell have been estimated. The dielectrophoretic spectrum of the real part of a complex function (Re[f ( ω] were computed using Visual Foxpro Version 6, which gave calculated values pertaining to electrical properties of the cell model as compared with experimental values. The process was repeated until the error percentile was in an acceptable range. The calculated parameters were the dielectric constants and the conductivities of the inner cytoplasm ( εic, σic, the outer cytoplasm ( εoc, σoc, the inner membrane ( εim, σim, the outer membrane ( εom, σom, the suspending solution( εs, σs and the thickness of each layer (dom, doc, dim, respectively. This computer program provides estimated values of cell electrical properties with high accuracy and required minimal computational time.

  4. 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

  5. Developing maintainability for tokamak fusion power systems. Phase I report. Volume I. Study results

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1977-10-01

    The overall purpose of the study is to identify design features of tokamak fusion power reactors which contribute to the achievement of high levels of maintainability. In this first phase, the principal emphasis is on scheduled maintenance whose frequency is determined by the life of the reactor first wall/blanket. Remote operations are baselined. Five conceptual reactor designs have been analyzed. Each concept is characterized by the size of the replaceable first wall/blanket module--large, intermediate, small--and whether access to the module was from the outside of the reactor, the inside of the reactor or a combination of both. The study results are expressed in terms of availability (scheduled maintenance downtime), the costs of maintenance (capital and recurring) and the percent effect of maintenance on the cost of electricity. During this first phase, the study benefitted significantly by the critical review of the feasibility of maintenance functions and the time-to-perform estimates by numerous persons involved in nuclear maintenance and remote operations

  6. Varennes Tokamak

    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

  7. Spherical galaxies.

    Science.gov (United States)

    Telles, J. E.; de Souza, R. E.; Penereiro, J. C.

    1990-11-01

    RESUMEN. Presentamos fotometria fotografica de 8 objetos y espectrosco- pla para 3 galaxias, las cuales son buenos candidatos para galaxias esfericas. Los resultados fotometricos se presentan en la forma de iso- fotas y de perfiles radiales promedlo, de los cuales se derivan para- metros estructurales. Estas observaciones combinadas con parametros di- namicos obtenidos de observaciones espectrosc6picas, son consistentes con el plano fundamental derivado por Djorgovski y Davis (1987). ABSTRACT. We present photographic surface photometry for 8 objects and spectroscopy for 3 galaxies which are good candidates for spherical galaxies. Photometric results are presented in the form of isophotes and mean radial profiles from which we derived structural parameters. These observations combined with dynamical parameters obtained from spectroscopic observations are consistent with the fundamental plane derived by Djorgovski and Davis (1987). Keq wo : CALAXIES-ELLIPTICAL

  8. Measurements of Prompt and MHD-Induced Fast Ion Loss from National Spherical Torus Experiment Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    D.S. Darrow; S.S. Medley; A.L. Roquemore; W.W. Heidbrink; A. Alekseyev; F.E. Cecil; J. Egedal; V.Ya. Goloborod' ko; N.N. Gorelenkov; M. Isobe; S. Kaye; M. Miah; F. Paoletti; M.H. Redi; S.N. Reznik; A. Rosenberg; R. White; D. Wyatt; V.A. Yavorskij

    2002-10-15

    A range of effects may make fast ion confinement in spherical tokamaks worse than in conventional aspect ratio tokamaks. Data from neutron detectors, a neutral particle analyzer, and a fast ion loss diagnostic on the National Spherical Torus Experiment (NSTX) indicate that neutral beam ion confinement is consistent with classical expectations in quiescent plasmas, within the {approx}25% errors of measurement. However, fast ion confinement in NSTX is frequently affected by magnetohydrodynamic (MHD) activity, and the effect of MHD can be quite strong.

  9. A study on the fusion reactor - Development of MHD stability and transport code for KT-2 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Koo; Shin, Kyo Jin [Pohang University of Science and Tecnology, Pohang (Korea, Republic of)

    1996-08-01

    MHD Stability analyses for KT-2 Tokamak were carried out by using CART (Resistive 3-D) Code. Linear Growth rates and linear perturbed eigen function of both N=0 axisymmetric mode and N=1 kink modes of highly elongated tokamak plasmas, in the presence of a conducting wall at various distances are computed and linear and nonlinear evolution of N=0 axisymmetric modes are simulated. 26 refs., 25 figs. (author)

  10. Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

    International Nuclear Information System (INIS)

    Kawashima, H.; Shimizu, K.; Takizuka, T.

    2007-01-01

    To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

  11. Development of hybrid frequency couplers for non-inductive current drive in a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berio, St.

    1996-11-04

    Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead of the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a non-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (author). 53 refs.

  12. Development of coupling systems at the hybrid frequency for the non-inductive current generation inside a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berio, S. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Aix-Marseille-1 Univ., 13 - Marseille (France)

    1996-12-31

    Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead to the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a on-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (authors) 53 refs.

  13. Spherical Torus Center Stack Design

    International Nuclear Information System (INIS)

    C. Neumeyer; P. Heitzenroeder; C. Kessel; M. Ono; M. Peng; J. Schmidt; R. Woolley; I. Zatz

    2002-01-01

    The low aspect ratio spherical torus (ST) configuration requires that the center stack design be optimized within a limited available space, using materials within their established allowables. This paper presents center stack design methods developed by the National Spherical Torus Experiment (NSTX) Project Team during the initial design of NSTX, and more recently for studies of a possible next-step ST (NSST) device

  14. The physics of tokamak start-up

    International Nuclear Information System (INIS)

    Mueller, D.

    2013-01-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current

  15. Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

    International Nuclear Information System (INIS)

    Zhang, Y. P.; Liu, Yi; Yuan, G. L.; Song, X. Y.; Yang, J. W.; Li, X.; Chen, W.; Li, Y.; Yan, L. W.; Song, X. M.; Yang, Q. W.; Duan, X. R.; Luo, X. B.; Liu, Y. Q.; Hua, Y.; Isobe, M.

    2014-01-01

    A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y. Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team , Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported

  16. Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Y. P., E-mail: zhangyp@swip.ac.cn; Liu, Yi; Yuan, G. L.; Song, X. Y.; Yang, J. W.; Li, X.; Chen, W.; Li, Y.; Yan, L. W.; Song, X. M.; Yang, Q. W.; Duan, X. R. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Luo, X. B.; Liu, Y. Q.; Hua, Y. [Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610041 (China); Isobe, M. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5259 (Japan)

    2014-05-15

    A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y. Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team , Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported.

  17. Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

    Science.gov (United States)

    Zhang, Y. P.; Liu, Yi; Luo, X. B.; Isobe, M.; Yuan, G. L.; Liu, Y. Q.; Hua, Y.; Song, X. Y.; Yang, J. W.; Li, X.; Chen, W.; Li, Y.; Yan, L. W.; Song, X. M.; Yang, Q. W.; Duan, X. R.

    2014-05-01

    A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y. Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team, Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported.

  18. 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

  19. Spherical torus, compact fusion at low field

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1985-02-01

    A spherical torus is obtained by retaining only the indispensable components on the inboard side of a tokamak plasma, such as a cooled, normal conductor that carries current to produce a toroidal magnetic field. The resulting device features an exceptionally small aspect ratio (ranging from below 2 to about 1.3), a naturally elongated D-shaped plasma cross section, and ramp-up of the plasma current primarily by noninductive means. As a result of the favorable dependence of the tokamak plasma behavior to decreasing aspect ratio, a spherical torus is projected to have small size, high beta, and modest field. Assuming Mirnov confinement scaling, an ignition spherical torus at a field of 2 T features a major radius of 1.5 m, a minor radius of 1.0 m, a plasma current of 14 MA, comparable toroidal and poloidal field coil currents, an average beta of 24%, and a fusion power of 50 MW. At 2 T, a Q = 1 spherical torus will have a major radius of 0.8 m, a minor radius of 0.5 m, and a fusion power of a few megawatts

  20. First results of spherical GEMs

    CERN Document Server

    Pinto, Serge Duarte; Brock, Ian; Croci, Gabriele; David, Eric; de Oliveira, Rui; Ropelewski, Leszek; van Stenis, Miranda; Taureg, Hans; Villa, Marco

    2010-01-01

    We developed a method to make GEM foils with a spherical geometry. Tests of this procedure and with the resulting spherical GEMs are presented. Together with a spherical drift electrode, a spherical conversion gap can be formed. This eliminates the parallax error for detection of x-rays, neutrons or UV photons when a gaseous converter is used. This parallax error limits the spatial resolution at wide scattering angles. Besides spherical GEMs, we have developed curved spacers to maintain accurate spacing, and a conical field cage to prevent edge distortion of the radial drift field up to the limit of the angular acceptance of the detector. With these components first tests are done in a setup with a spherical entrance window but a planar readout structure; results will be presented and discussed. A flat readout structure poses difficulties, however. Therefore we will show advanced plans to make a prototype of an entirely spherical double-GEM detector, including a spherical 2D readout structure. This detector w...

  1. Quantitative assessment and prediction of the contact area development during spherical tip indentation of glassy polymers.

    NARCIS (Netherlands)

    Pelletier, C.G.N.; Toonder, den J.M.J.; Govaert, L.E.; Hakiri, N.; Sakai, M.

    2008-01-01

    This paper describes the development of the contact area during indentation of polycarbonate. The contact area was measured in situ using an instrumented indentation microscope and compared with numerical simulations using an elasto-plastic constitutive model. The parameters in the model were

  2. 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

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

    International Nuclear Information System (INIS)

    Lee, V.D.

    1987-01-01

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

  4. Development of a second generation torsion balance based on a spherical superconducting suspension

    Science.gov (United States)

    Hammond, Giles D.; Speake, Clive C.; Matthews, Anthony J.; Rocco, Emanuele; Peña-Arellano, Fabian

    2008-02-01

    This paper describes the development of a second generation superconducting torsion balance to be used for a precision measurement of the Casimir force and a short range test of the inverse square law of gravity at 4.2K. The instrument utilizes niobium (Nb) as the superconducting element and employs passive damping of the parasitic modes of oscillation. Any contact potential difference between the torsion balance and its surroundings is nulled to within ≈50mV by applying known DC biases and fitting the resulting parabolic relationship between the measured torque and the applied voltage. A digital proportional-integral-derivative servo system has been developed and characterized in order to control the azimuthal position of the instrument. The angular acceleration and displacement noise are currently limited by the capacitive sensor at the level 3×10-8rads-2/√Hz and 30nm/√Hz at 100mHz. The possibility of lossy dielectric coatings on the surface of the torsion balance test masses is also investigated. Our measurements show that the loss angles δ are (1.5±2.3)×10-4 and (2.0±2.2)×10-4 at frequencies of 5 and 10mHz, respectively. These values of loss are not significant sources of error for measurements of the Casimir force using this experimental setup.

  5. Aquaporin 0 plays a pivotal role in refractive index gradient development in mammalian eye lens to prevent spherical aberration

    International Nuclear Information System (INIS)

    Kumari, S. Sindhu; Varadaraj, Kulandaiappan

    2014-01-01

    Highlights: • Intact AQP0 functions as fiber cell-to-fiber cell adhesion protein. • AQP0 facilitates reduction in extracellular space and lens water content. • AQP0 adhesion function aids in lens refractive index gradient (RING) formation. • AQP0 prevents lens spherical aberration by establishing RING. • AQP0 is critical for lens transparency and homeostasis. - Abstract: Aquaporin 0 (AQP0) is a transmembrane channel that constitutes ∼45% of the total membrane protein of the fiber cells in mammalian lens. It is critical for lens transparency and homeostasis as mutations and knockout cause autosomal dominant lens cataract. AQP0 functions as a water channel and as a cell-to-cell adhesion (CTCA) molecule in the lens. Our recent in vitro studies showed that the CTCA function of AQP0 could be crucial to establish lens refractive index gradient (RING). However, there is a lack of in vivo data to corroborate the role of AQP0 as a fiber CTCA molecule which is critical for creating lens RING. The present investigation is undertaken to gather in vivo evidence for the involvement of AQP0 in developing lens RING. Lenses of wild type (WT) mouse, AQP0 knockout (heterozygous, AQP0 +/− ) and AQP0 knockout lens transgenically expressing AQP1 (heterozygous AQP0 +/− /AQP1 +/− ) mouse models were used for the study. Data on AQP0 protein profile of intact and N- and/or C-terminal cleaved AQP0 in the lens by MALDI-TOF mass spectrometry and SDS–PAGE revealed that outer cortex fiber cells have only intact AQP0 of ∼28 kDa, inner cortical and outer nuclear fiber cells have both intact and cleaved forms, and inner nuclear fiber cells have only cleaved forms (∼26–24 kDa). Knocking out of 50% of AQP0 protein caused light scattering, spherical aberration (SA) and cataract. Restoring the lost fiber cell membrane water permeability (P f ) by transgene AQP1 did not reinstate complete lens transparency and the mouse lenses showed light scattering and SA. Transmission and

  6. Aquaporin 0 plays a pivotal role in refractive index gradient development in mammalian eye lens to prevent spherical aberration

    Energy Technology Data Exchange (ETDEWEB)

    Kumari, S. Sindhu [Physiology and Biophysics, Stony Brook University, Stony Brook, NY (United States); Varadaraj, Kulandaiappan, E-mail: kulandaiappan.varadaraj@stonybrook.edu [Physiology and Biophysics, Stony Brook University, Stony Brook, NY (United States); SUNY Eye Institute, New York, NY (United States)

    2014-10-03

    Highlights: • Intact AQP0 functions as fiber cell-to-fiber cell adhesion protein. • AQP0 facilitates reduction in extracellular space and lens water content. • AQP0 adhesion function aids in lens refractive index gradient (RING) formation. • AQP0 prevents lens spherical aberration by establishing RING. • AQP0 is critical for lens transparency and homeostasis. - Abstract: Aquaporin 0 (AQP0) is a transmembrane channel that constitutes ∼45% of the total membrane protein of the fiber cells in mammalian lens. It is critical for lens transparency and homeostasis as mutations and knockout cause autosomal dominant lens cataract. AQP0 functions as a water channel and as a cell-to-cell adhesion (CTCA) molecule in the lens. Our recent in vitro studies showed that the CTCA function of AQP0 could be crucial to establish lens refractive index gradient (RING). However, there is a lack of in vivo data to corroborate the role of AQP0 as a fiber CTCA molecule which is critical for creating lens RING. The present investigation is undertaken to gather in vivo evidence for the involvement of AQP0 in developing lens RING. Lenses of wild type (WT) mouse, AQP0 knockout (heterozygous, AQP0{sup +/−}) and AQP0 knockout lens transgenically expressing AQP1 (heterozygous AQP0{sup +/−}/AQP1{sup +/−}) mouse models were used for the study. Data on AQP0 protein profile of intact and N- and/or C-terminal cleaved AQP0 in the lens by MALDI-TOF mass spectrometry and SDS–PAGE revealed that outer cortex fiber cells have only intact AQP0 of ∼28 kDa, inner cortical and outer nuclear fiber cells have both intact and cleaved forms, and inner nuclear fiber cells have only cleaved forms (∼26–24 kDa). Knocking out of 50% of AQP0 protein caused light scattering, spherical aberration (SA) and cataract. Restoring the lost fiber cell membrane water permeability (P{sub f}) by transgene AQP1 did not reinstate complete lens transparency and the mouse lenses showed light scattering and SA

  7. 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)

  8. Development of new polysilsesquioxane spherical particles as stabilized active ingredients for sunscreens

    Science.gov (United States)

    Tolbert, Stephanie Helene

    Healthy skin is a sign of positive self-worth, attractiveness and vitality. Compromises to this are frequently caused by extended periods of recreation in the sun and in turn exposure to the harmful effects of UV radiation. To maintain strength and integrity, protection of the skin is paramount. This can be achieved by implementing skin-care products which contain sunscreen active ingredients that provide UV protection. Unfortunately, photo-degradation, toxicity, and photo-allergies limit the effectiveness of present day sunscreen ingredients. Currently, this is moderated by physically embedding within inert silica particles, but leaching of the active ingredient can occur, thereby negating protective efforts. Alternatively, this research details the preparation and investigation of bridged silsesquioxane analogues of commercial ingredients which can be chemically grafted to the silica matrix. Studies with bridged salicylate particles detail facile preparation, minimized leaching, and enhanced UV stability over physically encapsulated and pendant salicylate counterparts. In terms of UVB protective ability, the highest maintenance of sun protection factor (SPF) after extended UV exposure was achieved with bridged incorporation, and has been attributed to corollary UV stability. Additionally, bridged salicylate particles can be classified as broad-spectrum, and rate from moderate to good in terms of UVA protective ability. Particles incorporated with a bridged curcuminoid silsesquioxane were also prepared and displayed comparable results. As such, an attractive method for sunscreen isolation and stabilization has been developed to eliminate the problems associated with current sunscreens, all while maintaining the established UV absorbance profiles of the parent compound. To appreciate the technology utilized in this research, a thorough understanding of sol-gel science as it pertains to hybrid organic/silica particles, including methods of organic fragment

  9. Tokamak Systems Code

    International Nuclear Information System (INIS)

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

    1985-03-01

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

  10. 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

  11. 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

  12. 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 ...

  13. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    tokamak physics, from basic principles to interpretation of experimental data, and to a wider r eadership an elegant and authoritative introduction to the challenges that are associated with the development of the tokamak reactor, a source of limitless and clean thermonuclear power. This reference book should be on the shelf of every fusion scientist and graduate student. (book review)

  14. Development and integration of a 50 Hz pellet injection system for the Experimental Advanced Superconducting Tokamak (EAST)

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Xingjia [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Science Island Branch of Graduate School, University of Science and Technology of China, Hefei 230029 (China); Chen, Yue [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Jiansheng, E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Vinyar, Igor; Lukin, Alexander [PELIN, Saint-Petersburg (Russian Federation); Yuan, Xiaoling; Li, Changzheng; Liu, Haiqing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2017-01-15

    Highlights: • The design of the pumping system fits the operation requirement well not only theoretically but also experimentally. • The data showed that the averaged pellet injection velocity and propellant gas pressure had a relationship submitting to the power function. • The reliability of the injected pellet was mostly around 90% which is higher than the PI-20 system thanks to the improved pumping system and the new pellet fabrication and acceleration system. - Abstract: A 50 Hz pellet injection system, which is designed for edge-localized mode (ELM) control, has been successfully developed and integrated for the Experimental Advanced Superconducting Tokamak (EAST). Pellet injection is achieved by two separated injection system modules that can be operated independently from 1 to 25 Hz. The nominal injection velocity is 250 m/s with a scatter of ±50 m/s at a repetition rate of 50 Hz. A buffer tank and a two-stage differential pumping system of the pellet injection system was designed to increase hydrogen/deuterium ice quality and eliminate the influence of propellant gas on plasma operation, respectively. The pressure of the buffer tank could be pumped to 1 × 10{sup 2} Pa, and the pressure in the second differential chamber could reach 1 × 10{sup −4} Pa during the experiment. Engineering experiments, which consisted of 50 Hz pellet injection and guiding tube mock-up experiments, were also systematically carried out in a laboratory environment and demonstrated that the pellet injection system can reliably inject pellets at a repetitive frequency of 50 Hz.

  15. Spherical rhenium metal powder

    International Nuclear Information System (INIS)

    Leonhardt, T.; Moore, N.; Hamister, M.

    2001-01-01

    The development of a high-density, spherical rhenium powder (SReP) possessing excellent flow characteristics has enabled the use of advanced processing techniques for the manufacture of rhenium components. The techniques that were investigated were vacuum plasma spraying (VPS), direct-hot isostatic pressing (D-HIP), and various other traditional powder metallurgy processing methods of forming rhenium powder into near-net shaped components. The principal disadvantages of standard rhenium metal powder (RMP) for advanced consolidation applications include: poor flow characteristics; high oxygen content; and low and varying packing densities. SReP will lower costs, reduce processing times, and improve yields when manufacturing powder metallurgy rhenium components. The results of the powder characterization of spherical rhenium powder and the consolidation of the SReP are further discussed. (author)

  16. 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

  17. 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

  18. Development Of An Approach To Modeling Loading And Elution Of Spherical Resorcinol Formaldehyde Ion-Exchange Resin

    International Nuclear Information System (INIS)

    Aleman, S.; Hamm, L.; Smith, F.

    2011-01-01

    The current strategy for removal of cesium from the Hanford waste stream is ion-exchange using spherical Resorcinol-Formaldehyde (sRF) resin. The original resin of choice was granular SuperLig 644 resin and during testing of this resin several operational issues were identified. For example, the granular material had a high angle of internal friction resulting in fragmentation of resin particles along its edges during cycling and adverse hydraulic performance. Efforts to replace SuperLig 644 were undertaken and one candidate was the granular Resorcinol-Formaldehyde (RF) resin where experience with this cation exchanger dates back to the late 1940's. To minimize hydraulic concerns a spherical version of RF was developed and several different chemically produced batches were created. The 5E-370/641 batch of sRF was selected and for the last decade numerous studies have been performed (e.g., batch contact tests, column loading and elution tests). The Waste Treatment Plant (WTP) flowsheet shows that the aqueous phase waste stream will have a wide range of ionic concentrations (e.g., during the loading step 0-3 M free OH, 5+ M Na, 0-1 M K, 0-3 M NO 3 ). Several steps are required in the ion-exchange process to achieve the required Cs separation factors: loading, displacement, washing, elution, and regeneration. The sRF resin will be operated over a wide range in pH (i.e., pH of 12-14 during the loading step and pH of 0.01-1 during the elution step). During some of these steps very high levels of counter-ions and co-ions will be present within the aqueous phase. Alternative process feeds are under consideration as well (e.g., sodium levels as high as 8 M and column operation up to 45 C during loading, reduced and recycled HNO 3 during elution). In order to model the performance of sRF resin through an entire ion-exchange cycle, a more robust isotherm model is required. To achieve this more robust isotherm model requires knowledge of the numbers and kinds of fixed

  19. DEVELOPMENT OF AN APPROACH TO MODELING LOADING AND ELUTION OF SPHERICAL RESORCINOL FORMALDEHYDE ION-EXCHANGE RESIN

    Energy Technology Data Exchange (ETDEWEB)

    Aleman, S.; Hamm, L.; Smith, F.

    2011-10-03

    The current strategy for removal of cesium from the Hanford waste stream is ion-exchange using spherical Resorcinol-Formaldehyde (sRF) resin. The original resin of choice was granular SuperLig 644 resin and during testing of this resin several operational issues were identified. For example, the granular material had a high angle of internal friction resulting in fragmentation of resin particles along its edges during cycling and adverse hydraulic performance. Efforts to replace SuperLig 644 were undertaken and one candidate was the granular Resorcinol-Formaldehyde (RF) resin where experience with this cation exchanger dates back to the late 1940's. To minimize hydraulic concerns a spherical version of RF was developed and several different chemically produced batches were created. The 5E-370/641 batch of sRF was selected and for the last decade numerous studies have been performed (e.g., batch contact tests, column loading and elution tests). The Waste Treatment Plant (WTP) flowsheet shows that the aqueous phase waste stream will have a wide range of ionic concentrations (e.g., during the loading step 0-3 M free OH, 5+ M Na, 0-1 M K, 0-3 M NO{sub 3}). Several steps are required in the ion-exchange process to achieve the required Cs separation factors: loading, displacement, washing, elution, and regeneration. The sRF resin will be operated over a wide range in pH (i.e., pH of 12-14 during the loading step and pH of 0.01-1 during the elution step). During some of these steps very high levels of counter-ions and co-ions will be present within the aqueous phase. Alternative process feeds are under consideration as well (e.g., sodium levels as high as 8 M and column operation up to 45 C during loading, reduced and recycled HNO{sub 3} during elution). In order to model the performance of sRF resin through an entire ion-exchange cycle, a more robust isotherm model is required. To achieve this more robust isotherm model requires knowledge of the numbers and kinds of

  20. Development of Tokamak Reactor System Code and Performance for Early Realization of DEMO

    International Nuclear Information System (INIS)

    Hong, B. G.; Lee, D. W.; Kim, Y.

    2006-01-01

    To develop the concepts of DEMO and identify the design parameters, dependence on performance objectives, design features and physical and technical constraints have to be considered. System analyses are necessary to find device variables which optimize figures of merit such as major radius, ignition margin, divertor heat load, neutron wall load, etc. Demonstration fusion power plant, DEMO is regarded as the last step before the development of a commercial fusion reactor in Korea National Basic Plan for the Development of Fusion Energy. The DEMO should demonstrate a net electric power generation, a tritium self sufficiency, and the safety aspect of a power plant. Performance of DEMO for early realization has been investigated with a limited extension from the plasma physics and technology in the 2nd phase of the ITER operation (EPP phase)

  1. Global gyrokinetic simulation of tokamak transport

    International Nuclear Information System (INIS)

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

    1998-10-01

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

  2. Exhaust, ELM and Halo physics using the MAST tokamak

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. Development and Application of Predictive Tools for MHD Stability Limits in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Brennan, Dylan [Princeton Univ., NJ (United States); Miller, G. P. [Univ. of Tulsa, Tulsa, AZ (United States)

    2016-10-03

    This is a project to develop and apply analytic and computational tools to answer physics questions relevant to the onset of non-ideal magnetohydrodynamic (MHD) instabilities in toroidal magnetic confinement plasmas. The focused goal of the research is to develop predictive tools for these instabilities, including an inner layer solution algorithm, a resistive wall with control coils, and energetic particle effects. The production phase compares studies of instabilities in such systems using analytic techniques, PEST- III and NIMROD. Two important physics puzzles are targeted as guiding thrusts for the analyses. The first is to form an accurate description of the physics determining whether the resistive wall mode or a tearing mode will appear first as β is increased at low rotation and low error fields in DIII-D. The second is to understand the physical mechanism behind recent NIMROD results indicating strong damping and stabilization from energetic particle effects on linear resistive modes. The work seeks to develop a highly relevant predictive tool for ITER, advance the theoretical description of this physics in general, and analyze these instabilities in experiments such as ASDEX Upgrade, DIII-D, JET, JT-60U and NTSX. The awardee on this grant is the University of Tulsa. The research efforts are supervised principally by Dr. Brennan. Support is included for two graduate students, and a strong collaboration with Dr. John M. Finn of LANL. The work includes several ongoing collaborations with General Atomics, PPPL, and the NIMROD team, among others.

  4. Development status of the integrated tokamak simulator for K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kang, J. S.; Wang, J.; Hwang, Y. S. [Seoul National University, Seoul (Korea, Republic of); Jung, L. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korean fusion demonstration reactor (K-DEMO) study has been conducted to investigate the feasibility of an electricity generation, self-sustained tritium cycle, and component test facility. To estimate its capability, the integrated fusion operation simulator called INFRA has been developed by organizing relevant computational codes with standard data models and framework. The different modules of the integrated simulator are chosen among well-validated codes. Standard data models are directly linked with KSTAR experimental data so that the integrated simulator can be used for interpretative simulations but also for predictive simulations. In this study, the current status of code development and some examples of KSTAR interpretative simulations are reported. ITER integrated modelling and analysis suite is imported to K-DEMO data model to take over ITER experience and to accelerate collaboration with international IMAS community. Standardized rules and guideline have been developed by ITER team for many years. Based on strict policy, this data model has been established and updated. This data model is used for experimental and simulation results. The INFRA system has been utilized to be an alpha version of a KDEMO simulator. Database, framework, and module integration are conducted. A test equilibrium run for KSTAR is done by filling the database with experiment results. More modules will be incorporated in a near future. Validation with KSTAR data and benchmarking previous modelling activity is also planned in order to confirm the feasibility of this system.

  5. Development and Application of Predictive Tools for MHD Stability Limits in Tokamaks

    International Nuclear Information System (INIS)

    Brennan, Dylan; Miller, G. P.

    2016-01-01

    This is a project to develop and apply analytic and computational tools to answer physics questions relevant to the onset of non-ideal magnetohydrodynamic (MHD) instabilities in toroidal magnetic confinement plasmas. The focused goal of the research is to develop predictive tools for these instabilities, including an inner layer solution algorithm, a resistive wall with control coils, and energetic particle effects. The production phase compares studies of instabilities in such systems using analytic techniques, PEST- III and NIMROD. Two important physics puzzles are targeted as guiding thrusts for the analyses. The first is to form an accurate description of the physics determining whether the resistive wall mode or a tearing mode will appear first as β is increased at low rotation and low error fields in DIII-D. The second is to understand the physical mechanism behind recent NIMROD results indicating strong damping and stabilization from energetic particle effects on linear resistive modes. The work seeks to develop a highly relevant predictive tool for ITER, advance the theoretical description of this physics in general, and analyze these instabilities in experiments such as ASDEX Upgrade, DIII-D, JET, JT-60U and NTSX. The awardee on this grant is the University of Tulsa. The research efforts are supervised principally by Dr. Brennan. Support is included for two graduate students, and a strong collaboration with Dr. John M. Finn of LANL. The work includes several ongoing collaborations with General Atomics, PPPL, and the NIMROD team, among others.

  6. ITER tokamak device

    International Nuclear Information System (INIS)

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

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  7. Dust Measurements in Tokamaks

    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

  8. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

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

  9. Draining and drying process development of the Tokamak Cooling Water System of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seokho, E-mail: kims@ornl.gov [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Van Hove, Walter; Ferrada, Juan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Di Maio, Pietro Alessandro [University of Palermo, Viale delle Scienze, Palermo 90128 (Italy); Felde, David [Reactor and Nuclear Systems Division, ORNL, Oak Ridge, TN (United States); Raphael, Mitteau; Dell’Orco, Giovanni [ITER Organization, 13067 St Paul Lez Durance (France); Berry, Jan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2016-11-01

    Highlights: • A thermal-hydraulic model using RELAP was developed for the ITER FW/BLK modules to determine design parameters for the nitrogen blowout flow rate and pressure. • The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will sufficiently evacuate the water in blankets. • A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation. - Abstract: The ITER Organization (IO) developed a thermal-hydraulic (TH) model of the complex first wall and blanket (FW/BLK) cooling channels to determine gas flow rate and pressure required to effectively blow out the water in the FW/BLK. In addition, US ITER conducted experiments for selected geometries of FW/BLK flow channels to predict the blowout parameters. The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will ensure substantial evacuation of the water in blankets with just a few percent remaining in the blanket flow channels. A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation.

  10. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-22

    This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided.

  11. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1

    International Nuclear Information System (INIS)

    1995-01-01

    This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided

  12. Development of a Laser Ablation System Kit (LASK) for Tokamak in vessel tritium and dust inventory control

    International Nuclear Information System (INIS)

    Hernandez, C.; Roche, H.; Pocheau, C.; Grisolia, C.; Gargiulo, L.; Semerok, A.; Vatry, A.; Delaporte, P.; Mercadier, L.

    2009-01-01

    During Tokamak operation, Plasma Facing Components (PFCs) are subjected to severe interaction with plasma. As a consequence and independently of the PFCs composition, materials eroded and then re-deposited in the form of layers on the surfaces, can flake and produce dusts. These fragile structures are able to trap part of the hydrogenated species (tritium for example) in vessel inventory. For safety reasons, it is mandatory to measure and to control vessel dust and tritium inventory. Up to now, laser techniques are a part of the most promising methods able to solve these ITER open issues. Of special interest are laser systems loaded on a miniature tool that can be attached to a Multi Purpose Deployer (MPD) and used for laser treatments (detritiation and other), for PFCs chemical analysis as well as for micro particles recovery of dust produced during laser ablation. Such a system (Laser Ablation System Kit: LASK) is currently under development at IRFM and the following presentation will describe the current achievements of this project and the perspectives. In this paper, we will present an innovative compact system, which, loaded on a Multi Purpose Deployer, could allow operation in a harsh environment (pressure range from atmospheric to Ultra High Vacuum and temperature up to 120 deg. C). According to the process conditions, different treatments can be performed: at low laser fluence, PFCs thermal treatment will be expected, while at high laser fluence material will be ablated allowing Dust (and T) recovery as well as chemical analysis of material. This 'in-line' chemical analysis based on Laser Induced Breakdown Spectroscopy (LIBS) enables the ablation process to be controlled and preserves the substrate integrity. The paper will be focussed on the methodology followed during the LASK development and the method used to determine a laser process window able to remove co-deposited film without damaging the bulk material and taking into account external parameter

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  14. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

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

  15. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

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

  16. 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

  17. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

    The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

  18. Research and development of the JAERI large tokamak (JT-60), (4)

    International Nuclear Information System (INIS)

    Takashima, Tetsuo; Shimizu, Masatsugu; Ohta, Mitsuru; Minaguchi, Tadayoshi; Maeda, Hideto.

    1978-01-01

    A pair of fast-acting movable rail limiters are to be installed in the vacuum chamber of JT-60 to suppress skin current in the plasma column. They should travel across the vacuum chamber over a stroke of about 1 m in 0.1 sec in the build-up phase of the plasma current. Each movable limiter system consists of a drive mechanism, a vacuum seal, a bearing usable at high temperatures in a vacuum, a molybdenum rail limiter head and its auxiliary members. Various engineering problems are involved in constructing such a system because the design specifications outlined above exceed the present technology. A full-scale movable limiter, therefore, was designed, constructed and then put to mechanical, electrical and vacuum-technological tests. The model features a hydraulic drive mechanism with servovalves to control the oil flow. A special vacuum seal allowing a movement at high speeds was developed. It consists of welded bellows jointed together and connected to a pantograph at the joints. It allows uniform expansion of each bellows at high speeds. Molybdenum disulphide with 20% Ta is chosen as the most suitable bearing material after conducting tests on various bearing materials. The overall test of the model showed that its specifications were met with satisfactory reliability and reproducibility. Furthermore, the endurance test demonstrated that it functioned reliably over 50,000 times of operation. (author)

  19. 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)

  20. Tokamak concept innovations

    International Nuclear Information System (INIS)

    1986-04-01

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

  1. Overview of results from the National Spherical Torus Experiment (NSTX)

    Czech Academy of Sciences Publication Activity Database

    Gates, D.A.; Ahn, J.; Allain, J.; Andre, R.; Bastasz, R.; Bell, M.; Bell, R.; Belova, E.; Berkery, J.; Betti, R.; Bialek, J.; Biewer, T.; Bigelow, T.; Bitter, M.; Boedo, J.; Bonoli, P.; Boozer, A.; Brennan, D.; Breslau, J.; Brower, D.; Bush, C.; Canik, J.; Caravelli, G.; Carter, M.; Caughman, J.; Chang, C.; Crocker, N.; Darrow, D.; Delgado-Aparicio, L.; Diem, S.; D’Ippolito, D.; Domier, C.; Dorland, W.; Efthimion, P.; Ejiri, A.; Ershov, N.; Evans, T.; Feibush, E.; Fenstermacher, M.; Ferron, J.; Finkenthal, M.; Foley, J.; Frazin, R.; Fredrickson, E.; Fu, G.; Funaba, H.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Grisham, L.; Hahm, T.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hill, K.; Hillesheim, J.; Hillis, D.; Hirooka, Y.; Hosea, J.; Hu, B.; Humphreys, D.; Idehara, T.; Indireshkumar, K.; Ishida, A.; Jaeger, F.; Jarboe, T.; Jardin, S.; Jaworski, M.; Ji, H.; Kaita, R.; Kallman, J.; Katsuro-Hopkins, O.; Kawahata, K.; Kawamori, E.; Kaye, S.; Kessel, C.; Kimura, H.; Kolemen, E.; Krasheninnikov, H.; Krstic, P.; Ku, S.; Kubota, S.; Kugel, H.; La Haye, R.; Lao, L.; LeBlanc, B.; Lee, K.; Leuer, J.; Levinton, F.; Liang, Y.; Liu, D.; Luhmann Jr, N.; Maingi, R.; Majeski, R.; Manickam, J.; Mansfield, D.; Maqueda, R.; Mazzucato, E.; McCune, D.; McGeehan, B.; McKee, G.; Medley, S.; Menard, J.; Menon, M.; Meyer, H.; Mikkelsen, D.; Miloshevsky, G.; Mitarai, O.; Mueller, D.; Mueller, S.; Munsat, T.; Myra, J.; Nagayama, Y.; Nelson, B.; Nguyen, X.; Nishino, N.; Nishiura, M.; Nygren, R.; Ono, M.; Osborne, T.; Pacella, D.; Park, J.; Paul, S.; Peebles, W.; Penaflor, B.; Peng, M.; Phillips, C.; Pigarov, A.; Podesta, M.; Preinhaelter, Josef; Ram, A.; Raman, R.; Rasmussen, D.; Redd, A.; Reimerdes, H.; Rewoldt, G.; Ross, P.; Rowley, C.; Ruskov, E.; Russell, D.; Ruzic, D.; Ryan, P.; Sabbagh, S.; Schaffer, M.; Schuster, E.; Scott, S.; Shaing, K.; Sharpe, P.; Shevchenko, V.; Shinohara, K.; Sizyuk, V.; Skinner, C.; Smirnov, A.; Smith, D.; Snyder, P.; Solomon, W.; Sontag, A.; Soukhanovskii, V.; Stoltzfus-Dueck, T.; Stotler, D.; Strait, T.; Stratton, B.; Stutman, D.; Takahashi, R.; Takase, Y.; Tamura, N.; Tang, X.; Taylor, G.; Taylor, C.; Ticos, C.; Tritz, K.; Tsarouhas, D.; Turrnbull, A.; Tynan, G.; Ulrickson, M.; Umansky, M.; Urban, Jakub; Utergberg, E.; Walker, M.; Wampler, M.; Wang, J.; Wang, W.; Welander, A.; Whaley, J.; White, R.; Wilgen, J.; Wilson, R.; Wong, K.; Wright, J.; Xia, Z.; Xu, X.; Youchison, D.; Yu, G.; Yuh, H.; Zakharov, L.; Zemlyanov, D.; Zweben, S.; Choe, W.; Jung, H.; Kim, J.; Lee, W.; Park, H.

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104016-104016 ISSN 0029-5515. [IAEA Fusion Energy Conference/22nd./. Geneva, 13.10.2008-18.10.2008] R&D Projects: GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : NSTX * Spherical tokamaks * Overdense plasma * Conversion * Emission * Tokamaks * Elektron Bernstein waves Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://www.iop.org/EJ/article/0029-5515/49/10/104016/nf9_10_104016

  2. The Compact Ignition Tokamak

    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

  3. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

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

  6. Tokamak control simulator

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  7. 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)

  8. High Beta Tokamak research

    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

  9. Development of a six channel Fabry-Perot interferometer for continuous measurement of electron temperature of Tokamak plasma. Application to current diffusion study

    International Nuclear Information System (INIS)

    Talvard, M.

    1984-10-01

    It is shown how the properties of the electron cyclotron emission of a tokamak plasma can be used to measure the electron temperature. The design of a six channel Fabry-Perot interferometer is then described. This interferometer allows the measurement of the time evolution of the electron temperature profile of the plasma in the TFR tokamak. Using this technique interesting results have been obtained concerning the current penetration during the start up phase of a tokamak discharge [fr

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

    International Nuclear Information System (INIS)

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

    1992-03-01

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

  11. 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

  12. 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

  13. 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

  14. Current drive experiments in the HIT-II spherical tokamak

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Gu, P.; Isso, V.A.; Jewell, P.E.; McCollam, K.J.; Nelson, B.A.; Ramon, R.; Redd, A.J.; Sieck, P.E.; Smith, R.J.; Nagata, M.; Uyama, T.

    2001-01-01

    The Helicity Injected Torus (Hit) program has made progress in understanding relaxation and helicity injection current drive. Helicity-conserving MHD activity during the inductive (Ohmic) current ramp demonstrates the profile flattening needed for coaxial helicity injection (CHI). Results from cathode and anode central column (CC) CHI pulses are consistent with the electron locking model of current drive from a pure n=1 mode. Finally, low density CHI, compatible with Ohmic operation, has been achieved. Some enhancement of CHI discharges with the application of Ohmic is shown. (author)

  15. Design of the air-core transformer in spherical tokamak

    International Nuclear Information System (INIS)

    Wang Zhongtian; Jian Guangde; Li Fangzhu; Mao Guoping

    2002-01-01

    An ideal current distribution in the air-core transformer coils is obtained using variation principle. Climbing mountain method is utilized for optimizing the dimension and position of the real coils. Not only can the requirement of minimizing the stray field in the plasma region be guaranteed, but also integer turns for the coil can be realized. The latter brings a significant convenience to engineering

  16. Disruption mitigation studies on the Mega Amp Spherical Tokamak (MAST)

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, A.J., E-mail: at546@york.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Department of Physics, University of York, Helsington, York YO10 5DD (United Kingdom); Gibson, K.J. [Department of Physics, University of York, Helsington, York YO10 5DD (United Kingdom); Harrison, J.R. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Department of Physics, University of York, Helsington, York YO10 5DD (United Kingdom); Kirk, A. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lisgo, S.W. [ITER Organisation, Route de Vinon-sur-Verdon, St. Paul-lez-Durance, Cedex (France); Lehnen, M. [Institute for Energy Research - Plasma Physics, FZJ, Association EURATOM/FZJ, D-52425 Julich (Germany); Martin, R.; Naylor, G.; Scannell, R.; Cullen, A. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2011-08-01

    Disruptions pose a significant challenge in future devices where the increased stored energy can lead to unacceptably large transient heat loads on plasma facing components (PFCs). One means of mitigating disruptions is that of massive gas injection (MGI), which produces a radiative collapse of the plasma discharge through the injection of impurity gases. The MAST disruption mitigation system is capable of injecting up to 1.95 bar litres into the MAST vacuum vessel over a timescale of 1-2 ms, corresponding to a particle inventory of 5 x 10{sup 22}, around 100 times the plasma particle inventory. High speed infrared thermography, offering full divertor coverage, has shown a 60-70% reduction in divertor power loads during mitigation. A combination of high temporal (0.2 ms) and spatial resolution (1 cm) Thomson scattering and soft X-ray camera array data show evidence for a cooling front associated with the inward propagation of the injected impurities.

  17. Microtearing Instabilities and Electron Transport in the NSTX Spherical Tokamak

    International Nuclear Information System (INIS)

    Wong, K.L.; Kaye, S.; Mikkelsen, D.R.; Krommes, J.A.; Hill, K.; Bell, R.; LeBlanc, B.

    2007-01-01

    We report a successful quantitative account of the experimentally determined electron thermal conductivity χ e in a beam-heated H mode plasma by the magnetic fluctuations from microtearing instabilities. The calculated χ e based on existing nonlinear theory agrees with the result from transport analysis of the experimental data. Without using any adjustable parameter, the good agreement spans the entire region where there is a steep electron temperature gradient to drive the instability

  18. 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

  19. Numerical studies of edge localized instabilities in tokamaks

    International Nuclear Information System (INIS)

    Wilson, H.R.; Snyder, P.B.; Huysmans, G.T.A.; Miller, R.L.

    2002-01-01

    A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code

  20. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-08-01

    Dynamic control of the plasma position within the torus of a TOKAMAK fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. (auth)

  1. 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

  2. Tokamak ARC damage

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

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

  4. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

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

  5. Development of a hydrothermal method to synthesize spherical ZnSe nanoparticles: Appropriate templates for hollow nanostructures

    Directory of Open Access Journals (Sweden)

    S. Gharibe

    2014-01-01

    Full Text Available Hydrothermal method was used to synthesize pure ZnSe nanosphere materials. The effects of the reducing agent amount, the reaction time and temperature were investigated on the purity of ZnSe. Also, the effects of surfactants such as sodium dodecyl sulfate (SDS (anionic and cetyl trimethylammonium bromide (CTAB (cationic were studied on the morphology of ZnSe. The prepared nanospheres were characterized using XRD, SEM, TEM and UV-Vis spectroscopy. Through these techniques, it was found that the pure ZnSe nanoparticles have a zinc blend structure and in a spherical form with average diameter of 30 nm. DOI: http://dx.doi.org/10.4314/bcse.v28i1.5

  6. FY1995 development of artificial arm 'SMART ARM' by spherical ultrasonic motor; 1995 nendo kyumen choonpa motor wo mochiita jinko gishu smart arm no kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    The project has an intention of development of new type artificial arm by spherical ultrasonic motor. We have succeeded in developing new type of spherical ultrasonic motor with three DOF. And we have succeeded in applying the motor to an artificial arm. This arm have advantages of small size, low weight torque comparing with conventional ones. We demonstrated them the new arm behaved well and it had good controlabilty. (NEDO)

  7. Role of symmetry-breaking induced by Er × B shear flows on developing residual stresses and intrinsic rotation in the TEXTOR tokamak

    International Nuclear Information System (INIS)

    Xu, Y.; Shesterikov, I.; Berte, M.; Dumortier, P.; Van Schoor, M.; Vergote, M.; Hidalgo, C.; Krämer-Flecken, A.; Koslowski, R.

    2013-01-01

    Direct measurements of residual stress (force) have been executed at the edge of the TEXTOR tokamak using multitip Langmuir and Mach probes, together with counter-current NBI torque to balance the existing toroidal rotation. Substantial residual stress and force have been observed at the plasma boundary, confirming the existence of a finite residual stress as possible mechanisms to drive the intrinsic toroidal rotation. In low-density discharges, the residual stress displays a quasi-linear dependence on the local pressure gradient, consistent with theoretical predictions. At high-density shots the residual stress and torque are strongly suppressed. The results show close correlation between the residual stress and the E r × B flow shear rate, suggesting a minimum threshold of the E × B flow shear required for the k ∥ symmetry breaking. These findings provide the first experimental evidence of the role of E r × B sheared flows in the development of residual stresses and intrinsic rotation. (letter)

  8. Specific features of the occurrence, development, and re-compaction of spall and shear fractures in spherically-convergent shells made of unalloyed iron and some steels under their spherical explosive loading

    International Nuclear Information System (INIS)

    Kozlov, E.A.; Brichikov, S.A.; Gorbachev, D.M.; Brodova, I.G.; Yablonskikh, T.I.

    2007-01-01

    Results of comparative metallographic examination of recovered shells exposed to explosive loading in two modes (with and without a heavy casing confining explosion products scatter) are presented. The shells were made of high-purity and technical-grade unalloyed iron with the initial grain size 250 and 125 μm, steel 30KhGSA in delivery state and quenched up to HR C 35...40, austenitic stainless steel 12Kh18N10T. The heavy casing used in experiments is demonstrated to ensure a rather compact convergence of shells destroyed at high radii. In the described comparative experiments, one managed to compile the 12Kh18N10T steel shell, after it was spalled at high radii and exposed to shear fracture and spallation layer fragmentation at medium radii, into a compact sphere but failed to do the same with the 30KhGSA quenched steel shell after it was fractured according to spall and shear mechanisms at high and medium radii. Polar zones of this steel shell have obvious undercompressed areas due to significant dissipative losses to overcome the shear strength. Occurrence, development, and re-compaction of spall and shear fractures in spherically-convergent shells made of materials, which were already carefully investigated in 1D- and 2D-geometry experiments, were systematically studied in order to verify and validate new physical models of dynamic fractures, as well as up-to-date used in 1D-, 2D- and 3D-numerical algorithms [ru

  9. Are Nanoparticles Spherical or Quasi-Spherical?

    Science.gov (United States)

    Sokolov, Stanislav V; Batchelor-McAuley, Christopher; Tschulik, Kristina; Fletcher, Stephen; Compton, Richard G

    2015-07-20

    The geometry of quasi-spherical nanoparticles is investigated. The combination of SEM imaging and electrochemical nano-impact experiments is demonstrated to allow sizing and characterization of the geometry of single silver nanoparticles. © 2015 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim.

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

    International Nuclear Information System (INIS)

    Lopez C, R.

    1986-07-01

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

  11. 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

  12. Spherical neutron generator

    Science.gov (United States)

    Leung, Ka-Ngo

    2006-11-21

    A spherical neutron generator is formed with a small spherical target and a spherical shell RF-driven plasma ion source surrounding the target. A deuterium (or deuterium and tritium) ion plasma is produced by RF excitation in the plasma ion source using an RF antenna. The plasma generation region is a spherical shell between an outer chamber and an inner extraction electrode. A spherical neutron generating target is at the center of the chamber and is biased negatively with respect to the extraction electrode which contains many holes. Ions passing through the holes in the extraction electrode are focused onto the target which produces neutrons by D-D or D-T reactions.

  13. Next Step Spherical Torus Design Studies

    International Nuclear Information System (INIS)

    Neumeyer, C.; Heitzenroeder, P.; Kessel, C.; Ono, M.; Peng, M.; Schmidt, J.; Woolley, R.; Zatz, I.

    2002-01-01

    Studies are underway to identify and characterize a design point for a Next Step Spherical Torus (NSST) experiment. This would be a ''Proof of Performance'' device which would follow and build upon the successes of the National Spherical Torus Experiment (NSTX) a ''Proof of Principle'' device which has operated at PPPL since 1999. With the Decontamination and Decommissioning (DandD) of the Tokamak Fusion Test Reactor (TFTR) nearly completed, the TFTR test cell and facility will soon be available for a device such as NSST. By utilizing the TFTR test cell, NSST can be constructed for a relatively low cost on a short time scale. In addition, while furthering spherical torus (ST) research, this device could achieve modest fusion power gain for short-pulse lengths, a significant step toward future large burning plasma devices now under discussion in the fusion community. The selected design point is Q=2 at HH=1.4, P subscript ''fusion''=60 MW, 5 second pulse, with R subscript ''0''=1.5 m, A=1.6, I subscript ''p''=10vMA, B subscript ''t''=2.6 T, CS flux=16 weber. Most of the research would be conducted in D-D, with a limited D-T campaign during the last years of the program

  14. Current drive by spheromak injection into a tokamak

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  15. 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)

  16. 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

  17. Preliminary project of s Thomson scattering system for the ETE tokamak; Projeto preliminar de um sistema de espalhamento Thomson para o Tokamak ETE

    Energy Technology Data Exchange (ETDEWEB)

    Berni, Luiz Angelo

    1997-12-31

    This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light 4 refs., 26 figs.

  18. Magnet design considerations for Tokamak fusion reactors

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  19. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). However, the results of this investigation are believed to be generally applicable to the broad class of the next generation of experimental tokamak facilities such as ETF. The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties are compared to the benefits and conclusions and recommendations are developed on resolving the issue

  20. 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)

  1. 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

  2. Mechanisms of Stochastic Diffusion of Energetic Ions in Spherical Tori

    Energy Technology Data Exchange (ETDEWEB)

    Ya.I. Kolesnichenko; R.B. White; Yu.V. Yakovenko

    2001-01-18

    Stochastic diffusion of the energetic ions in spherical tori is considered. The following issues are addressed: (I) Goldston-White-Boozer diffusion in a rippled field; (ii) cyclotron-resonance-induced diffusion caused by the ripple; (iii) effects of non-conservation of the magnetic moment in an axisymmetric field. It is found that the stochastic diffusion in spherical tori with a weak magnetic field has a number of peculiarities in comparison with conventional tokamaks; in particular, it is characterized by an increased role of mechanisms associated with non-conservation of the particle magnetic moment. It is concluded that in current experiments on National Spherical Torus eXperiment (NSTX) the stochastic diffusion does not have a considerable influence on the confinement of energetic ions.

  3. 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

  4. Development of a flexible Doppler reflectometry system and its application to turbulence characterization in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Troester, Carolin Helma

    2008-04-15

    An essential challenge in present fusion plasma research is the study of plasma turbulence. The turbulence behavior is investigated experimentally on the ASDEX Upgrade tokamak using Doppler reflectometry, a diagnostic technique sensitive to density fluctuations at a specific wavenumber k {sub perpendicular} {sub to}. This microwave radar diagnostic utilizes localized Bragg backscattering of the launched beam (k{sub 0}) by the density fluctuations at the plasma cutoff layer. The incident angle {theta} selects the probed k {sub perpendicular} {sub to} via the Bragg condition k {sub perpendicular} {sub to} {approx} 2k{sub 0}sin{theta}. The measured Doppler shifted frequency spectrum allows the determination of the perpendicular plasma rotation velocity, u {sub perpendicular} {sub to} =v{sub E} {sub x} {sub B}+v{sub turb}, directly from the Doppler frequency shift(f{sub D} = u {sub perpendicular} {sub to} k {sub perpendicular} {sub to} /2{pi}), and the turbulence amplitude from the backscattered power level. This thesis work presents a survey of u {sub perpendicular} {sub to} radial profiles and k {sub perpendicular} {sub to} spectrum measurements for a variety of plasma conditions obtained by scanning the antenna tilt angle. This was achieved by extending the existing V-band Doppler reflectometry system (50 - 75 GHz) with a new W-band system (75 - 110 GHz), which was especially designed for measuring the k {sub perpendicular} {sub to} spectrum and additionally expands the radial coverage into the plasma core region. It consists of a remote steerable antenna with an adjustable line of sight allowing for dynamic wavenumber selection up to 25 cm {sup -1} and a reflectometer with a 'phase locked loop' stabilized transmitter allowing for the precise determination of the instrument response function. The proper system functionality was demonstrated by laboratory testing and benckmarking against the V-band system. The new profile measurements obtained show a

  5. Ripple induced trapped particle loss in tokamaks

    International Nuclear Information System (INIS)

    White, R.B.

    1996-05-01

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

  6. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

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

  7. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-04-01

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

  8. 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

  9. 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

  10. A Fast Shutdown Technique for Large Tokamaks

    International Nuclear Information System (INIS)

    Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.

    1999-01-01

    A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR

  11. Electronic system of TBR tokamak device

    International Nuclear Information System (INIS)

    Silva, R.P. da.

    1980-01-01

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

  12. Overview of Tokamak Results

    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

  13. Effects of enhanced elongation and paramagnetism on the parameter space of the ignition spherical torus

    International Nuclear Information System (INIS)

    Strickler, D.J.; Peng, Y-K.M.; Borowski, S.K.; Selcow, E.C.; Miller, J.B.

    1985-01-01

    The Ignition Spherical Torus (IST) is a small aspect ratio device retaining only indispensable components along the major axis of a tokamak plasma, such as a cooled, normal conductor producing a toroidal magnetic field. The IST is expected to be a cost-effective approach to ignition by taking advantage of low field, large natural plasma elongation, high plasma current, high beta, and tokamak confinement. These result in compact, high-performance devices with relatively simple magnetic systems as compared with ignition tokamaks of larger aspect ratio. The plasma enhancement of the toroidal field on axis, or plasma paramagnetism, is significant in the IST. The use of this plasma-enhanced field in conventional tokamak beta and density limits leads to increased plasma pressure and performance and therefore smaller device size for a given ignition margin

  14. The tokamak hybrid reactor

    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.)

  15. Ion diagnostics for the tokamak boundary

    International Nuclear Information System (INIS)

    Matthews, G.F.

    1991-01-01

    In this paper, recent developments in ion diagnostic probes for tokamak boundary plasmas are discussed. Three areas are covered: retarding field analysers, sniffer probes and plasma ion mass spectrometers. The contribution of these diagnostics to our understanding of plasma surface interactions is summarised. (author)

  16. Development of nanostructured porous TiO2 thick film with uniform spherical particles by a new polymeric gel process for dye-sensitized solar cell applications

    International Nuclear Information System (INIS)

    Bakhshayesh, A.M.; Mohammadi, M.R.

    2013-01-01

    A novel simple synthetic procedure for fabrication of high surface area nanostructured TiO 2 electrode with uniform particles for photovoltaic application is reported. Modifying the TiO 2 particulate sol by pH adjustment together with employment of a polymeric agent, so-called polymeric gel process, was developed. The polymeric gel process was used to deposit nanostructured thick electrode by dip coating incorporated in dye-sensitized solar cells (DSSCs). X-ray diffraction (XRD) analysis revealed that deposited film was composed of primary nanoparticles with average crystallite size in the range 21-39 nm. Field emission scanning electron microscope (FE-SEM) images showed that deposited film had nanostructured and porous morphology containing uniform spherical particles with diameter about 2.5 μm. The spherical particles were made of small nanoparticles with average grain size of 60 nm improving light scattering and dye loading of the DSSC. Moreover, atomic force microscope (AFM) analysis verified that the roughness mean square of prepared electrode was low, enhancing electron transport to the counter electrode. Photovoltaic measurements showed that solar cell made of polymeric gel process had higher photovoltaic performance than that made of conventional paste. An enhancement of power conversion efficiency from 4.54%, for conventional paste, to 6.21%, for polymeric gel process, was achieved. Electrochemical impedance spectroscopy (EIS) study showed that the recombination process in solar cell made of polymeric gel process was slower than that in solar cell made of conventional paste. The presented strategy would open up new insight into fabrication of low-cost TiO 2 DSSCs with high power conversion efficiency

  17. 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

  18. Preliminary project of s Thomson scattering system for the ETE tokamak

    International Nuclear Information System (INIS)

    Berni, Luiz Angelo

    1997-01-01

    This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light

  19. High beta tokamaks

    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

  20. 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

  1. Reconnection in tokamaks

    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

  2. Research using small tokamaks

    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

  3. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

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

  4. Compact tokamak reactors

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Texas Experimental Tokamak

    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

  6. The spherical harmonics method, II (application to problems with plane and spherical symmetry)

    Energy Technology Data Exchange (ETDEWEB)

    Mark, C

    1958-12-15

    The application of the spherical harmonic method to problems with plane or spherical symmetry is discussed in detail. The numerical results of some applications already made are included to indicate the degree of convergence obtained. Formulae for dealing with distributions of isotropic sources are developed. Tables useful in applying the method are given in Section 11. (author)

  7. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

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

  8. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

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

  9. Physics Basis for a Spherical Torus Power Plant

    International Nuclear Information System (INIS)

    Kessel, C.E.; Menard, J.; Jardin, S.C.; Mau, T.K.

    1999-01-01

    The spherical torus, or low-aspect-ratio tokamak, is considered as the basis for a fusion power plant. A special class of wall-stabilized high-beta high-bootstrap fraction low-aspect-ratio tokamak equilibrium are analyzed with respect to MHD stability, bootstrap current and external current drive, poloidal field system requirements, power and particle exhaust and plasma operating regime. Overall systems optimization leads to a choice of aspect ratio A = 1:6, plasma elongation kappa = 3:4, and triangularity delta = 0:64. The design value for the plasma toroidal beta is 50%, corresponding to beta N = 7:4, which is 10% below the ideal stability limit. The bootstrap fraction of 99% greatly alleviates the current drive requirements, which are met by tangential neutral beam injection. The design is such that 45% of the thermal power is radiated in the plasma by Bremsstrahlung and trace Krypton, with Neon in the scrapeoff layer radiating the remainder

  10. Whist code calculations of ignition margin in an ignition tokamak

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    A simple global model was developed to determine the ignition margin of tokamaks including electron and ion conduction losses. A comparison of this model with results from a 1 1/2 dimensional Whist code is made

  11. FY 2006 Miniature Spherical Retroreflectors Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Anheier, Norman C.; Bernacki, Bruce E.; Krishnaswami, Kannan

    2006-12-28

    Research done by the Infrared Photonics team at Pacific Northwest National Laboratory (PNNL) is focused on developing miniature spherical retroreflectors using the unique optical and material properties of chalcogenide glass to reduce both performance limiting spherical aberrations. The optimized optical performance will provide efficient signal retroreflection that enables a broad range of remote detection scenarios for mid-wave infrared (MWIR) and long-wave infrared (LWIR) sensing applications. Miniature spherical retroreflectors can be developed to aid in the detection of signatures of nuclear proliferation or other chemical vapor or radiation signatures. Miniature spherical retroreflectors are not only well suited to traditional LIDAR methods for chemical plume detection and identification, but could enable remote detection of difficult semi-volatile chemical materials or low level radiation sources.

  12. Annual report of Division of Thermonuclear Fusion Research and Division of Large Tokamak Development for the period of April 1, 1976 to March 31, 1977

    International Nuclear Information System (INIS)

    1978-02-01

    Research and development activities in the two divisions are closely related. 1) Theoretical and computational studies continued on tokamak confinement and heating related to experimental problems. Studies on NBI heating in JT-60 were completed. 2) Experimental studies on impurities, density control and effects of density fluctuations were made in JFT-2. Neutral beams up to 30 keV and 8 A were injected into JFT-2 plasma perpendicularly. The ion temperature was increased by 10% - 15%, which is in agreement with the prediction by classical Fokker-Planck theory. In JFT-2a(DIVA), plasma-wall interaction (behavior of heavy and light impurities) was studies. The divertor of DIVA reduced the plasma-wall interaction and hence the radiation loss due to heavy impurities by a factor of 3. A grazing-incidence vacuum monochromator was first used in impurity studies in JFT-2 and JFT-2a. 3) Technological improvements were made raising efficiencies of operation, maintenance and plasma research. 4) Neutral beam injector test stand ITS-2 of 100 keV was completed. Construction of a 200 kW, 650 MHz radiofrequency heating system for JFT-2 was started. 5) Sputterings of molybdenum and pyrolytic graphite by low-energy protons and chemical reaction rates of pyrolytic graphite with protons were measured. Honeycomb structure greatly reduced the sputtered particles. 6) The superconducting magnet development group made the design of cluster test apparatus and the development of large current superconductor. 7) Phase-I preliminary design of experimental fusion reactor JXFR was completed and preliminary safety evaluation of JXFR was made. 8) Detailed design of JT-60 was completed in November 1976. Engineering development contracts were all completed by March 1977. 9) Engineering studies and tests on critical components of JT-4 with non-circular plasma cross section and divertors were made, after the preliminary design in fiscal year 1975. (auth.)

  13. Effects of isotropic alpha populations on tokamak ballooning stability

    International Nuclear Information System (INIS)

    Spong, D.A.; Sigmar, D.J.; Tsang, K.T.; Ramos, J.J.; Hastings, D.E.; Cooper, W.A.

    1986-12-01

    Fusion product alpha populations can significantly influence tokamak stability due to coupling between the trapped alpha precessional drift and the kinetic ballooning mode frequency. Careful, quantitative evaluations of these effects are necessary in burning plasma devices such as the Tokamak Fusion Test Reactor and the Joint European Torus, and we have continued systematic development of such a kinetic stability model. In this model we have considered a range of different forms for the alpha distribution function and the tokamak equilibrium. Both Maxwellian and slowing-down models have been used for the alpha energy dependence while deeply trapped and, more recently, isotropic pitch angle dependences have been examined

  14. 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

  15. Transient electromagnetic analysis in tokamaks using TYPHOON code

    International Nuclear Information System (INIS)

    Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

    1996-01-01

    The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

  16. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1983-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

  17. Magnetic Diagnostics for Equilibrium Reconstructions in the Presence of Nonaxisymmetric Eddy Current Distributions in Tokamaks

    International Nuclear Information System (INIS)

    Kaita, R.; Kozub, T.; Logan, N.; Majeski, R.; Menard, J.; Zakharov, L.

    2010-01-01

    The lithium tokamak experiment (LTX) is a modest-sized spherical tokamak (R 0 = 0.4 m and a = 0.26 m) designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 C. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.

  18. 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

  19. PPPL Laboratory Program Development Activities for fiscal year 1993

    International Nuclear Information System (INIS)

    1993-01-01

    This report discusses the following topics: Advanced Tokamak Studies; Princeton Spherical Tokamak Experiment; Medium-Scale Long-Pulse Device Study; Collaborations Planning and Exploration; Divertor Simulator Studies; Gyrofluid Simulation; Feedback Kink Study; Stellarator Studies; High-Field Magnet Studies; Analysis of Helically Wound Solenoids; X-Ray Lithography with Tokamak Radiation; Magnetospheric Plasma Circulation; and Projection Lithography with X-Ray Laser

  20. The Spherical Deformation Model

    DEFF Research Database (Denmark)

    Hobolth, Asgar

    2003-01-01

    Miller et al. (1994) describe a model for representing spatial objects with no obvious landmarks. Each object is represented by a global translation and a normal deformation of a sphere. The normal deformation is defined via the orthonormal spherical-harmonic basis. In this paper we analyse the s...

  1. 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

  2. Tokamak reactor startup power

    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

  3. 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.)

  4. Theory of tokamak plasmas

    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.

  5. 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

  6. Compact magnetic confinement fusion: Spherical torus and compact torus

    Directory of Open Access Journals (Sweden)

    Zhe Gao

    2016-05-01

    Full Text Available The spherical torus (ST and compact torus (CT are two kinds of alternative magnetic confinement fusion concepts with compact geometry. The ST is actually a sub-category of tokamak with a low aspect ratio; while the CT is a toroidal magnetic configuration with a simply-connected geometry including spheromak and field reversed pinch. The ST and CT have potential advantages for ultimate fusion reactor; while at present they can also provide unique fusion science and technology contributions for mainstream fusion research. However, some critical scientific and technology issues should be extensively investigated.

  7. 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

  8. Friction factor for water flow through packed beds of spherical and non-spherical particles

    Directory of Open Access Journals (Sweden)

    Kaluđerović-Radoičić Tatjana

    2017-01-01

    Full Text Available The aim of this work was the experimental evaluation of different friction factor correlations for water flow through packed beds of spherical and non-spherical particles at ambient temperature. The experiments were performed by measuring the pressure drop across the bed. Packed beds made of monosized glass spherical particles of seven different diameters were used, as well as beds made of 16 fractions of quartz filtration sand obtained by sieving (polydisperse non-spherical particles. The range of bed voidages was 0.359–0.486, while the range of bed particle Reynolds numbers was from 0.3 to 286 for spherical particles and from 0.1 to 50 for non-spherical particles. The obtained results were compared using a number of available literature correlations. In order to improve the correlation results for spherical particles, a new simple equation was proposed in the form of Ergun’s equation, with modified coefficients. The new correlation had a mean absolute deviation between experimental and calculated values of pressure drop of 9.04%. For non-spherical quartz filtration sand particles the best fit was obtained using Ergun’s equation, with a mean absolute deviation of 10.36%. Surface-volume diameter (dSV necessary for correlating the data for filtration sand particles was calculated based on correlations for dV = f(dm and Ψ = f(dm. [Project of the Serbian Ministry of Education, Science and Technological Development, Grant no. ON172022

  9. Density limits in Tokamaks

    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)

  10. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.

    1984-05-01

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

  11. 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

  12. Tokamak pump limiters

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  13. 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

  14. Overview of physics results from the conclusive operation of the National Spherical Torus Experiment

    Czech Academy of Sciences Publication Activity Database

    Sabbagh, S.A.; Ahn, J-W.; Allain, J.; Andre, R.; Balbaky, A.; Bastasz, R.; Battaglia, D.; Bell, M.; Bell, R.; Beiersdorfer, P.; Belova, E.; Berkery, J.; Betti, R.; Bialek, J.; Bigelow, T.; Bitter, M.; Boedo, J.; Bonoli, P.; Boozer, A.; Bortolon, A.; Boyle, D.; Brennan, D.; Breslau, J.; Buttery, R.; Canik, J.; Caravelli, G.; Chang, C.; Crocker, N.; Darrow, D.; Davis, B.; Delgado-Aparicio, L.; Diallo, A.; Ding, S.; D’Ippolito, D.; Domier, C.; Dorland, W.; Ethier, S.; Evans, T.; Ferron, J.; Finkenthal, M.; Foley, J.; Fonck, R.; Frazin, R.; Fredrickson, E.; Fu, G.; Gates, D.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Gray, T.; Guo, Y.; Guttenfelder, W.; Hahm, T.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hill, K.; Hirooka, Y.; Hooper, E.B.; Hosea, J.; Jardin, S.; Jaworski, M.; Kaita, R.; Kallman, J.; Katsuro-Hopkins, O.; Kaye, S.; Kessel, C.; Kim, J.; Kolemen, E.; Kramer, G.; Krasheninnikov, S.; Kubota, S.; Kugel, H.; La Haye, R.J.; Lao, L.; LeBlanc, B.; Lee, W.; Lee, K.; Leuer, J.; Levinton, F.; Liang, Y.; Liu, D.; Lore, J.; Luhmann Jr, N.; Maingi, R.; Majeski, R.; Manickam, J.; Mansfield, D.; Maqueda, R.; McKee, G.; Medley, S.; Meier, E.; Menard, J.; Menon, M.; Meyer, H.; Mikkelsen, D.; Miloshevsky, G.; Mueller, D.; Munsat, T.; Myra, J.; Nelson, B.; Nishino, N.; Nygren, R.; Ono, M.; Osborne, T.; Park, J.; Park, Y.S.; Paul, S.; Peebles, W.; Penaflor, B.; Perkins, R.J.; Phillips, C.; Pigarov, A.; Podesta, M.; Preinhaelter, Josef; Raman, R.; Ren, Y.; Rewoldt, G.; Rognlien, T.; Ross, P.; Rowley, C.; Ruskov, E.; Russell, D.; Ruzic, D.; Ryan, P.; Schaffer, M.; Schuster, E.; Scotti, F.; Shaing, K.; Shevchenko, V.; Shinohara, K.; Sizyuk, V.; Skinner, C.H.; Smirnov, A.; Smith, A.; Snyder, P.; Solomon, W.; Sontag, A.; Soukhanovskii, V.; Stoltzfus-Dueck, T.; Stotler, D.; Stratton, B.; Stutman, D.; Takahashi, H.; Takase, Y.; Tamura, N.; Tang, X.; Taylor, G.; Taylor, C.; Tritz, K.; Tsarouhas, D.; Umansky, M.; Urban, Jakub; Untergberg, E.; Walker, M.; Wampler, W.; Wang, W.; Whaley, J.; White, R.; Wilgen, J.; Wilson, R.; Wong, K.L.; Wright, J.; Xia, Z.; Youchison, D.; Yu, G.; Yuh, H.; Zakharov, L.; Zemlyanov, D.; Zimmer, G.; Zweben, S.J.

    2013-01-01

    Roč. 53, č. 10 (2013), s. 104007-104007 ISSN 0029-5515. [IAEA Fusion Energy Conference/24./. San Diego, 08.10.2012-13.10.2012] R&D Projects: GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : NSTX * Spherical tokamaks * Overdense plasma * Conversion * Emission * Tokamaks * Electron Bernstein waves Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://iopscience.iop.org/0029-5515/53/10/104007/pdf/0029-5515_53_10_104007.pdf

  15. Development of a dry-mechanical graphite separation process and elimination of the separated carbon for the reprocessing of spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Kronschnabel, H.

    1982-01-01

    Due to the C-14 distribution the separation of the particle-free outer region of the spherical HTR fuel element with subsequent solidification of the separated carbon makes it possible to reduce by half the remaining C-14 inventory in the inner particle region to be further treated. Separation of the particle-free outer region by a newly developed sphere-peeling milling machine, conditioning the graphite into compacts and in-situ cementation into a salt-mine are the basic elements of this head-end process variation. An annual cavern volume of approx. 2000 m 3 will be needed to ultimately store the graphite of the particle-free outer region, which corresponds to a reprocessing capacity of 50 GWsub(e) installed HTR power. The brush-disintegration of the remaining inner particle region and the resulting peel-brush-preparation are capable of separating 95% of the graphite without any heavy metal losses. With the mentioned reprocessing capacity an annual cavern volume of approx. 16.500 m 3 is required. (orig.) [de

  16. Opto-mechanical design and development status of an all spherical five lenses focal reducer for the 2.3 m Thai National Telescope

    Science.gov (United States)

    Buisset, Christophe; Prasit, Apirat; Lépine, Thierry; Poshyachinda, Saran; Soonthornthum, Boonrucksar; Deboos, Alexis

    2016-07-01

    The National Astronomical Research Institute (NARIT) is currently developing an all spherical five lenses focal reducer to image a FOV circular of diameter Δθ = 14.6' on the 4K camera with a pixel scale equal to 0.42''/pixel. The spatial resolution will be better than 1.2'' over the full visible spectral domain [400 nm, 800 nm]. The relative irradiance between the ghost and the science images will be lower than 10-4. The maximum distortion will be lower than 1% and the maximum angle of incidence on the filters will be equal to 8°. The focal reducer comprises 1 doublet L1 located at the fork entrance and 1 triplet L2 located in front of the camera. The doublet L1 will be mounted on a tip-tilt mount placed on a robotic sliding rail. L1 will thus be placed in the optical path during the observations with the 4K camera and will be removed during the observations with the other instruments. The triplet L2 will be installed on the instrument cube in front of the camera equipped with the filter wheel. The glass will be manufactured in a specialized company, the mechanical parts will be manufactured by using the NARIT Computer Numerical Control machine and the lenses will be integrated at NARIT. In this paper, we describe the optical and mechanical designs and we present the geometrical performance, the transmission budget and the results of the stray light analyses.

  17. A programmatic framework for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.

    1993-01-01

    Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program

  18. Study of a spherical torus based volumetric neutron source for nuclear technology testing and development. Final report of a scientific research supported by the USDOE/SBIR program

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1999-01-01

    A plasma based, deuterium and tritium (DT) fueled, volumetric 14 MeV neutron source (VNS) has been considered as a possible facility to support the development of the demonstration fusion power reactor (DEMO). It can be used to test and develop necessary fusion blanket and divertor components and provide sufficient database, particularly on the reliability of nuclear components necessary for DEMO. The VNS device complement to ITER by reducing the cost and risk in the development of DEMO. A low cost, scientifically attractive, and technologically feasible volumetric neutron source based on the spherical torus (ST) concept has been conceived. The ST-VNS, which has a major radius of 1.07 m, aspect ratio 1.4, and plasma elongation 3, can produce a neutron wall loading from 0.5 to 5 MW/m 2 at the outboard test section with a modest fusion power level from 38 to 380 MW. It can be used to test necessary nuclear technologies for fusion power reactor and develop fusion core components include divertor, first wall, and power blanket. Using staged operation leading to high neutron wall loading and optimistic availability, a neutron fluence of more than 30 MW-y/m 2 is obtainable within 20 years of operation. This will permit the assessments of lifetime and reliability of promising fusion core components in a reactor relevant environment. A full scale demonstration of power reactor fusion core components is also made possible because of the high neutron wall loading capability. Tritium breeding in such a full scale demonstration can be very useful to ensure the self-sufficiency of fuel cycle for a candidate power blanket concept

  19. Design studies of Tokamak power reactor in JAERI

    International Nuclear Information System (INIS)

    Tone, T.; Nishikawa, M.; Tanaka, Y.

    1985-01-01

    Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

  20. Automated Fault Detection for DIII-D Tokamak Experiments

    International Nuclear Information System (INIS)

    Walker, M.L.; Scoville, J.T.; Johnson, R.D.; Hyatt, A.W.; Lee, J.

    1999-01-01

    An automated fault detection software system has been developed and was used during 1999 DIII-D plasma operations. The Fault Identification and Communication System (FICS) executes automatically after every plasma discharge to check dozens of subsystems for proper operation and communicates the test results to the tokamak operator. This system is now used routinely during DIII-D operations and has led to an increase in tokamak productivity

  1. Facility approach to tokamak operation

    International Nuclear Information System (INIS)

    Edmonds, P.H.; Gabbard, W.A.

    1981-01-01

    In anticipation of the appearance of more advanced tokamaks and other fusion relevant experiments, program has been established at ORNL to systemically identify the requirements of an effective machine operations group. This program is presently applied to the ISX-B experiment. With its continuing development, it is expected to provide major support in the identification of potential problem areas and to assist in the generation of the necessary procedures for forthcoming devices. The present and future generations of large plasma devices will function as facilities, operated by an operations group as service to the plasma physicists and diagnosticians. The purpose of the program discussed here is to develop and to encourage an orderly transition to the facility-like style of operation

  2. Simulation models for tokamak plasmas

    International Nuclear Information System (INIS)

    Dimits, A.M.; Cohen, B.I.

    1992-01-01

    Two developments in the nonlinear simulation of tokamak plasmas are described: (A) Simulation algorithms that use quasiballooning coordinates have been implemented in a 3D fluid code and a 3D partially linearized (Δf) particle code. In quasiballooning coordinates, one of the coordinate directions is closely aligned with that of the magnetic field, allowing both optimal use of the grid resolution for structures highly elongated along the magnetic field as well as implementation of the correct periodicity conditions with no discontinuities in the toroidal direction. (B) Progress on the implementation of a likeparticle collision operator suitable for use in partially linearized particle codes is reported. The binary collision approach is shown to be unusable for this purpose. The algorithm under development is a complete version of the test-particle plus source-field approach that was suggested and partially implemented by Xu and Rosenbluth

  3. Spherical proton emitters

    International Nuclear Information System (INIS)

    Berg, S.; Semmes, P.B.; Nazarewicz, W.

    1997-01-01

    Various theoretical approaches to proton emission from spherical nuclei are investigated, and it is found that all the methods employed give very similar results. The calculated decay widths are found to be qualitatively insensitive to the parameters of the proton-nucleus potential, i.e., changing the potential parameters over a fairly large range typically changes the decay width by no more than a factor of ∼3. Proton half-lives of observed heavy proton emitters are, in general, well reproduced by spherical calculations with the spectroscopic factors calculated in the independent quasiparticle approximation. The quantitative agreement with experimental data obtained in our study requires that the parameters of the proton-nucleus potential be chosen carefully. It also suggests that deformed proton emitters will provide invaluable spectroscopic information on the angular momentum decomposition of single-proton orbitals in deformed nuclei. copyright 1997 The American Physical Society

  4. 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

  5. Plea for stellarator funding raps tokamaks

    International Nuclear Information System (INIS)

    Blake, M.

    1992-01-01

    The funding crunch in magnetic confinement fusion development has moved the editor of a largely technical publication to speak out on a policy issue. James A. Rome, who edits Stellarator News from the Fusion Energy Division at Oak Ridge National Laboratory, wrote an editorial that appeared on the front page of the May 1992 issue. It was titled open-quotes The US Stellarator Program: A Time for Renewal,close quotes and while it focused chiefly on that subject (and lamented the lack of funding for the operation of the existing ATF stellarator at Oak Ridge), it also cited some of the problems inherent in the mainline MCF approach--the tokamak--and stated that if the money can be found for further tokamak design upgrades, it should also be found for stellarators. Rome wrote, open-quotes There is growing recognition in the US, and elsewhere, that the conventional tokamak does not extrapolate to a commercially competitive energy source except with very high field coils ( 1000 MWe).close quotes He pointed up open-quotes the difficulty of simultaneously satisfying conflicting tokamak requirements for efficient current drive, high bootstrap-current fraction, complete avoidance of disruptions, adequate beta limits, and edge-plasma properties compatible with improved (H-mode) confinement and acceptable erosion of divertor plates.close quotes He then called for support for the stellarator as open-quotes the only concept that has performance comparable to that achieved in tokamaks without the plasma-current-related limitations listed above.close quotes

  6. The Spherical Deformation Model

    DEFF Research Database (Denmark)

    Hobolth, Asgar

    2003-01-01

    Miller et al. (1994) describe a model for representing spatial objects with no obvious landmarks. Each object is represented by a global translation and a normal deformation of a sphere. The normal deformation is defined via the orthonormal spherical-harmonic basis. In this paper we analyse the s...... a single central section of the object. We use maximum-likelihood-based inference for this purpose and demonstrate the suggested methods on real data....

  7. Cooperative effects in spherical spasers

    DEFF Research Database (Denmark)

    Bordo, Vladimir

    2017-01-01

    A fully analytical semiclassical theory of cooperative optical processes which occur in an ensemble of molecules embedded in a spherical core-shell nanoparticle is developed from first principles. Both the plasmonic Dicke effect and spaser generation are investigated for the designs in which...... a shell/core contains an arbitrarily large number of active molecules in the vicinity of a metallic core/shell. An essential aspect of the theory is an ab initio account of the feedback from the core/shell boundaries which significantly modifies the molecular dynamics. The theory provides rigorous, albeit...

  8. Software development and its description for Geoid determination based on Spherical-Cap-Harmonics Modelling using digital-zenith camera and gravimetric measurements hybrid data

    Science.gov (United States)

    Morozova, K.; Jaeger, R.; Balodis, J.; Kaminskis, J.

    2017-10-01

    Over several years the Institute of Geodesy and Geoinformatics (GGI) was engaged in the design and development of a digital zenith camera. At the moment the camera developments are finished and tests by field measurements are done. In order to check these data and to use them for geoid model determination DFHRS (Digital Finite element Height reference surface (HRS)) v4.3. software is used. It is based on parametric modelling of the HRS as a continous polynomial surface. The HRS, providing the local Geoid height N, is a necessary geodetic infrastructure for a GNSS-based determination of physcial heights H from ellipsoidal GNSS heights h, by H=h-N. The research and this publication is dealing with the inclusion of the data of observed vertical deflections from digital zenith camera into the mathematical model of the DFHRS approach and software v4.3. A first target was to test out and validate the mathematical model and software, using additionally real data of the above mentioned zenith camera observations of deflections of the vertical. A second concern of the research was to analyze the results and the improvement of the Latvian quasi-geoid computation compared to the previous version HRS computed without zenith camera based deflections of the vertical. The further development of the mathematical model and software concerns the use of spherical-cap-harmonics as the designed carrier function for the DFHRS v.5. It enables - in the sense of the strict integrated geodesy approach, holding also for geodetic network adjustment - both a full gravity field and a geoid and quasi-geoid determination. In addition, it allows the inclusion of gravimetric measurements, together with deflections of the vertical from digital-zenith cameras, and all other types of observations. The theoretical description of the updated version of DFHRS software and methods are discussed in this publication.

  9. Recent results from the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Maingi, R; Bell, M G; Bell, R E; Bialek, J; Bourdelle, C; Bush, C E; Darrow, D S; Fredrickson, E D; Gates, D A; Gilmore, M; Gray, T; Jarboe, T R; Johnson, D W; Kaita, R; Kaye, S M; Kubota, S; Kugel, H W; LeBlanc, B P; Maqueda, R J; Mastrovito, D; Medley, S S; Menard, J E; Mueller, D; Nelson, B A; Ono, M; Paoletti, F; Park, H K; Paul, S F; Peebles, T; Peng, Y-K M; Phillips, C K; Raman, R; Rosenberg, A L; Roquemore, A L; Ryan, P M; Sabbagh, S A; Skinner, C H; Soukhanovskii, V A; Stutman, D; Swain, D W; Synakowski, E J; Taylor, G; Wilgen, J; Wilson, J R; Wurden, G A; Zweben, S J

    2003-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect-ratio fusion research facility whose research goal is to make a determination of the attractiveness of the spherical torus concept in the areas of high-β stability, confinement, current drive, and divertor physics. Remarkable progress was made in extending the operational regime of the device in FY 2002. In brief, β t of 34% and β N of 6.5 were achieved. H-mode became the main operational regime, and energy confinement exceeded conventional aspect-ratio tokamak scalings. Heating was demonstrated with the radiofrequency antenna, and signatures of current drive were observed. Current initiation with coaxial helicity injection produced discharges of 400 kA, and first measurements of divertor heat flux profiles in H-mode were made

  10. Improvement of tokamak performance by injection of electrons

    International Nuclear Information System (INIS)

    Ono, Masayuki.

    1992-12-01

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

  11. The scientific program of the Tokamak de Varennes

    International Nuclear Information System (INIS)

    Daughney, C.C.

    1989-01-01

    The Tokamak de Varennes (TdeV) is the principal research tool of the Centre canadien de fusion magnetique (CCFM). This article places the Tokamak de Varennes within the framework of the Canadian National Fusion Program (NFP) and describes the scientific program of the TdeV as it was presented at the April 1989 meeting of the CCFM Advisory Committee. The CCFM scientific plant contains three main elements: tokamak development, research on transport and equilibrium in plasmas, and research on the plasma-wall problem. Phase I of the experimental program, commissioning the tokamak and the diagnostic systems, has been completed. Phase II of the experimental program will begin in December 1989 with the plasma boundary defined by a magnetic divertor and the power supplies and vacuum system capable of creating a sequence of one-second plasma pulses. (3 figs., 3 refs.) (L.L.)

  12. 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

  13. The density limit in Tokamaks

    International Nuclear Information System (INIS)

    Alladio, F.

    1985-01-01

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

  14. Burn Control Mechanisms in Tokamaks

    Science.gov (United States)

    Hill, M. A.; Stacey, W. M.

    2015-11-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

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

    International Nuclear Information System (INIS)

    1977-03-01

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

  16. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 29, Analysis results. Volume 3

    International Nuclear Information System (INIS)

    Xu, Minfeng

    1995-01-01

    The electromagnetic analysis is mainly based on model built with 3-D electromagnetic software OPERA/TOSCA. In the process of evaluating the software package, some models are also built with 3-D boundary element electromagnetic software AMPERES. Fortran programs are also developed at B ampersand W to perform Monte-Carlo simulations of the field error analysis to assist tolerance determinations

  17. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 21, Materials and processes selection. Volume 2

    International Nuclear Information System (INIS)

    Smith, B.R.

    1995-01-01

    This document identifies the candidate materials and manufacturing processes selected for development of the TPX Toroidal Field (TF) Magnet. Supporting rationale and selection criteria are provided for justification and the materials properties database report is included for completeness. Specific properties for each material selection are included in this document

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

    Directory of Open Access Journals (Sweden)

    Mitsuru Kikuchi

    2010-11-01

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

  19. Fundamentals of spherical array processing

    CERN Document Server

    Rafaely, Boaz

    2015-01-01

    This book provides a comprehensive introduction to the theory and practice of spherical microphone arrays. It is written for graduate students, researchers and engineers who work with spherical microphone arrays in a wide range of applications.   The first two chapters provide the reader with the necessary mathematical and physical background, including an introduction to the spherical Fourier transform and the formulation of plane-wave sound fields in the spherical harmonic domain. The third chapter covers the theory of spatial sampling, employed when selecting the positions of microphones to sample sound pressure functions in space. Subsequent chapters present various spherical array configurations, including the popular rigid-sphere-based configuration. Beamforming (spatial filtering) in the spherical harmonics domain, including axis-symmetric beamforming, and the performance measures of directivity index and white noise gain are introduced, and a range of optimal beamformers for spherical arrays, includi...

  20. 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

  1. 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.

  2. Development of large high-voltage pressure insulators for the Princeton TFTR [Tokamak Fusion Test Reactor] flexible transmission lines

    International Nuclear Information System (INIS)

    Scalise, D.T.; Fong, E.; Haughian, J.; Prechter, R.

    1986-10-01

    Specially formulated insulator materials with improved strength and high-voltage properties were developed and used for critical components of the flexible transmission lines to the TFTR neutral beam ion sources. These critical components are plates which support central conductors as they exit the high-voltage power supply and enter the ion source enclosure. Each plate acts both as a high-voltage insulator and as a pressure barrier to the SF 6 insulating gas. The original plate was made of commercial glass-epoxy laminate which limited the plate voltage capacity. The newly developed insulator is made of specially-formulated cycloalphatic Di-epoxide whose isotropic properties exhibit increased arc resistance. It is cast in one piece with skirts which greatly increase the breakdown voltage. This paper discusses the design, fabrication and testing of the new insulator

  3. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-01-01

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

  4. Lower hybrid current drive in shaped tokamaks

    International Nuclear Information System (INIS)

    Kesner, J.

    1993-01-01

    A time dependent lower hybrid current drive tokamak simulation code has been developed. This code combines the BALDUR tokamak simulation code and the Bonoli/Englade lower hybrid current drive code and permits the study of the interaction of lower hybrid current drive with neutral beam heating in shaped cross-section plasmas. The code is time dependent and includes the beam driven and bootstrap currents in addition to the current driven by the lower hybrid system. Examples of simulations are shown for the PBX-M experiment which include the effect of cross section shaping on current drive, ballooning mode stabilization by current profile control and sawtooth stabilization. A critical question in current drive calculations is the radial transport of the energetic electrons. The authors have developed a response function technique to calculate radial transport in the presence of an electric field. The consequences of the combined influences of radial diffusion and electric field acceleration are discussed

  5. Module description of TOKAMAK equilibrium code MEUDAS

    International Nuclear Information System (INIS)

    Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa

    2002-01-01

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

  6. Initial plasma production by induction electric field on QUEST tokamak

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke

    2007-01-01

    Induction electric field by center solenoid coil plays a roll to produce initial plasma. According to Townsend avalanche theory, minimum electric field for plasma breakdown depends on neutral gas pressure and connection length. On QUEST spherical tokamak, a connection length is evaluated as 966m on null point neighborhood with coil current ratio I PF26 /I CS =0.1, and induction electric field considering eddy current of vacuum vessel is evaluated as about 0.1 V/m on null point neighborhood. With Townsend avalanche theory, these values manage to produce initial plasma on QUEST. (author)

  7. 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

  8. 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)

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

    International Nuclear Information System (INIS)

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

    1982-06-01

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

  10. Design constraints on magnet systems of future tokamaks based on experiences of present s.c. magnet development

    International Nuclear Information System (INIS)

    Heinz, W.; Jeske, U.; Komarek, P.; Krauth, H.

    1983-01-01

    In view of the urgent need for superconductivity in the next generation of big fusion devices and the identified gap between aimed data and the state of the art, impressive development programs are running world-wide, e.g. the IEA-Large Coil Task (LCT) and magnets for near term experiments (T15, Tore Supra). During the development work for all these magnet systems and simultaneously running design studies, especially the INTOR-study, some critical problem areas, e.g. concerning NbTi-conductor design and manufacturing and coil fabrication could be solved, others like the limitations by fatigue stresses for coil case and support structure turned out to be more stringent than anticipated. This paper tries to show which plasma physics parameters place especially severe constraints to magnet design, like PF-pulse number and amplitude at the TF-coils, so that they should be chosen with strongest care. It further points out which technologies under these circumstances are still missing or unproven with respect to the INTOR-like generation of fusion experiments. Further effort is mainly required for fatigue load behaviour of materials and components, high field windings and poloidal field coils. (author)

  11. ECH system developments including the design of an intelligent fault processor on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Ponce, D.; Lohr, J.; Tooker, J.F.; O'Neill, R.C.; Moeller, C.P.; Doane, J.L.; Noraky, S.; Dubovenko, K.; Gorelov, Y.A.; Cengher, M.; Penaflor, B.G.; Ellis, R.A.

    2011-01-01

    A new generation fault processor is in development which is intended to increase fault handling flexibility and reduce the number of incomplete DIII-D shots due to gyrotron faults. The processor, which is based upon a field programmable gate array device, will analyze signals for aberrant operation and ramp down high voltage to try to avoid hard faults. The processor will then attempt to ramp back up to an attainable operating point. The new generation fault processor will be developed during an expansion of the electron cyclotron heating (ECH) areas that will include the installation of a depressed collector gyrotron and associated equipment. Existing systems will also be upgraded. Testing of real-time control of the ECH launcher poloidal drives by the DIII-D plasma control system will be completed. The ECH control system software will be upgraded for increased scalability and to increase operator productivity. Resources permitting, all systems will receive an extra layer of interlocks for the filament and magnet power supplies, added shielding for the tank electronics, programmable filament boost shape for long pulses, and electronics upgrades for the installation of the advanced fault processor.

  12. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties to the benefits and conclusions and recommendations on resolving the issue are discussed

  13. Tokamak with liquid metal toroidal field coil

    International Nuclear Information System (INIS)

    Ohkawa, T.; Schaffer, M.J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof

  14. Physics parameter space of tokamak ignition devices

    International Nuclear Information System (INIS)

    Selcow, E.C.; Peng, Y.K.M.; Uckan, N.A.; Houlberg, W.A.

    1985-01-01

    This paper describes the results of a study to explore the physics parameter space of tokamak ignition experiments. A new physics systems code has been developed to perform the study. This code performs a global plasma analysis using steady-state, two-fluid, energy-transport models. In this paper, we discuss the models used in the code and their application to the analysis of compact ignition experiments. 8 refs., 8 figs., 1 tab

  15. A novel approach to linearization of the electromagnetic parameters of tokamaks with an iron core

    Energy Technology Data Exchange (ETDEWEB)

    Fu, P. E-mail: fupeng@mail.ipp.ac.cn; Liu, Z.Z.; Zou, J.H

    2002-05-01

    The equivalent model of an iron core tokamak is developed, in which the electromagnetic parameters of several pairs of coils in opposite series (PCOS) are not dependent on the saturation of the iron core during tokamak operation. With this the electromagnetic parameters of all the coils in an iron core tokamak can be linearized, As an example, the electromagnetic parameters of Hefei Super-conductive Tokamak with iron core (HT-7) are linearized, and it is in good agreement with the experimental results. The linearization approach can be applied in real time plasma control and electromagnetic analysis.

  16. Application and Continued Development of Thin Faraday Collectors as a Lost Ion Diagnostic for Tokamak Fusion Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    F. Ed Cecil

    2011-06-30

    This report summarizes the accomplishment of sixteen years of work toward the development of thin foil Faraday collectors as a lost energetic ion diagnostic for high temperature magnetic confinement fusion plasmas. Following initial, proof of principle accelerator based studies, devices have been tested on TFTR, NSTX, ALCATOR, DIII-D, and JET (KA-1 and KA-2). The reference numbers refer to the attached list of publications. The JET diagnostic KA-2 continues in operation and hopefully will provide valuable diagnostic information during a possible d-t campaign on JET in the coming years. A thin Faraday foil spectrometer, by virtue of its radiation hardness, may likewise provide a solution to the very challenging problem of lost alpha particle measurements on ITER and other future burning plasma machines.

  17. Wave Driven Fast Ion Loss in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; Cheng, C.Z.; Darrow, D.; Fu, G.; Gorelenkov, N.N.; Kramer, G.; Medley, S.S.; Menard, J.; Roquemore, L.; Stutman, D.; White, R.B.

    2003-01-01

    The study of fast ion instabilities in conventional aspect ratio tokamaks is motivated in large part by their potential to negatively impact the ignition threshold in fusion reactors by causing fast ion losses. Spherical tokamak's (ST), with intrinsically low magnetic fields, are particularly susceptible to fast ion driven instabilities. The 3.5 MeV alpha's from the D-T [deuterium-tritium] fusion reaction in proposed ST reactors will have velocities much higher than the Alfven speed. The Larmor radius of the fusion alphas, normalized to the plasma size, will also be larger than for conventional aspect ratio tokamak reactors. The resulting longer wavelengths of the *AE instabilities will be more effective in driving fast ion loss. The change in magnetic topology also influences the mode structure, as in the case of the Compressional Alfven Eigenmodes (CAE) seen on NSTX

  18. Dust limit management strategy in tokamaks

    International Nuclear Information System (INIS)

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Delaporte, P.; Douai, D.; Garnier, D.; Gauthier, E.; Gulden, W.; Hong, S.H.; Pitcher, S.; Rodriguez, L.; Taylor, N.; Tesini, A.; Vartanian, S.; Vatry, A.; Wykes, M.

    2009-01-01

    Dust is produced in tokamaks by the interaction between the plasma and the plasma facing components. Dust has not yet been of a major concern in existing tokamaks mainly because the quantity is small and these devices are not nuclear facilities. However, in ITER and in future reactors, it will represent operational and potential safety issues. From a safety point of view, in order to control the potential dust hazard, the current ITER strategy is based on a defense in depth approach designed to provide reliable confinement systems, to avoid failures, and to measure and minimise the dust inventory. In addition, R and D is put in place for optimisation of the proposed methods, such as improvement of measurement, dust cleaning and the reduction of dust production. The aim of this paper is to present the approach for the control of the dust inventory, relying on the monitoring of envelope values and the development of removal techniques already developed in the existing tokamaks or plasma dedicated devices or which will need further research and development in order to be integrated in ITER.

  19. Dust limit management strategy in tokamaks

    Science.gov (United States)

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Delaporte, P.; Douai, D.; Garnier, D.; Gauthier, E.; Gulden, W.; Hong, S. H.; Pitcher, S.; Rodriguez, L.; Taylor, N.; Tesini, A.; Vartanian, S.; Vatry, A.; Wykes, M.

    2009-06-01

    Dust is produced in tokamaks by the interaction between the plasma and the plasma facing components. Dust has not yet been of a major concern in existing tokamaks mainly because the quantity is small and these devices are not nuclear facilities. However, in ITER and in future reactors, it will represent operational and potential safety issues. From a safety point of view, in order to control the potential dust hazard, the current ITER strategy is based on a defense in depth approach designed to provide reliable confinement systems, to avoid failures, and to measure and minimise the dust inventory. In addition, R&D is put in place for optimisation of the proposed methods, such as improvement of measurement, dust cleaning and the reduction of dust production. The aim of this paper is to present the approach for the control of the dust inventory, relying on the monitoring of envelope values and the development of removal techniques already developed in the existing tokamaks or plasma dedicated devices or which will need further research and development in order to be integrated in ITER.

  20. Holographic Spherically Symmetric Metrics

    Science.gov (United States)

    Petri, Michael

    The holographic principle (HP) conjectures, that the maximum number of degrees of freedom of any realistic physical system is proportional to the system's boundary area. The HP has its roots in the study of black holes. It has recently been applied to cosmological solutions. In this article we apply the HP to spherically symmetric static space-times. We find that any regular spherically symmetric object saturating the HP is subject to tight constraints on the (interior) metric, energy-density, temperature and entropy-density. Whenever gravity can be described by a metric theory, gravity is macroscopically scale invariant and the laws of thermodynamics hold locally and globally, the (interior) metric of a regular holographic object is uniquely determined up to a constant factor and the interior matter-state must follow well defined scaling relations. When the metric theory of gravity is general relativity, the interior matter has an overall string equation of state (EOS) and a unique total energy-density. Thus the holographic metric derived in this article can serve as simple interior 4D realization of Mathur's string fuzzball proposal. Some properties of the holographic metric and its possible experimental verification are discussed. The geodesics of the holographic metric describe an isotropically expanding (or contracting) universe with a nearly homogeneous matter-distribution within the local Hubble volume. Due to the overall string EOS the active gravitational mass-density is zero, resulting in a coasting expansion with Ht = 1, which is compatible with the recent GRB-data.

  1. Topology of tokamak orbits

    International Nuclear Information System (INIS)

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

    1978-09-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  3. Start of the international tokamak physics activity

    International Nuclear Information System (INIS)

    Campbell, D.

    2001-01-01

    This newsletter comprises a summary on the start of the International Tokamak Physics activity (ITPA) by Dr. D. Campbell, Chair of the ITPA Co-ordinating Committee. As the ITER EDA drew to a close, it became clear that it was desirable to establish a new mechanism in order to promote the continued development of the physics basis for burning plasma experiments and to preserve the invaluable collaborations between the major international fusion communities which had been established through the ITER physics expert groups. As a result of the discussions of the representatives of the European Union, Japan, the Russian Federation and the United States the agreed principles for conducting the International Tokamak Physics Activity (ITPA) were elaborated and ITPA topical physics groups were organized

  4. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1981-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and USSR. The Zero-Phase of the INTOR Workshop, which was conducted during 1979, assessed the technical data base that would support the construction of the next major device in the tokamak program to operate in the early 1990s and defined the objectives and characteristics of this device. The INTOR workshop was extended into phase-1, the Definition Phase, in early 1980. The objective of the Phase-1 Workshop was to develop a conceptual design of the INTOR experiment. The purpose of this paper is to give an overview of the work of the Phase-1 INTOR Workshop (January 1980-June 1981, with emphasis upon the conceptual design

  5. Magnetohydrodynamic stability of tokamak edge plasmas

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  6. 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

  7. 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.

  8. Comparison between stellarator and tokamak divertor transport

    International Nuclear Information System (INIS)

    Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

    2010-11-01

    The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

  9. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1996-06-01

    A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).

  10. Design of the ITER tokamak assembly tools

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyunki [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)], E-mail: hkpark@nfri.re.kr; Lee, Jaehyuk; Kim, Taehyung [SFA Engineering Corp., 42-7 Palyong-dong, Changwon-si, Gyeongsangnam-do 641-847 (Korea, Republic of); Song, Yunju [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of); Im, Kihak [ITER Organization, CEA Cadarasche, 13108 Saint Paul-lez-Durance (France); Kim, Byungchul; Lee, Hyeongon; Jung, Ki-Jung [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)

    2008-12-15

    ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40 deg. sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values.

  11. Liquid tin limiter for FTU tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  12. Coherent structures in tokamak plasmas workshop: Proceedings

    International Nuclear Information System (INIS)

    Koniges, A.E.; Craddock, G.G.

    1992-08-01

    Coherent structures have the potential to impact a variety of theoretical and experimental aspects of tokamak plasma confinement. This includes the basic processes controlling plasma transport, propagation and efficiency of external mechanisms such as wave heating and the accuracy of plasma diagnostics. While the role of coherent structures in fluid dynamics is better understood, this is a new topic for consideration by plasma physicists. This informal workshop arose out of the need to identify the magnitude of structures in tokamaks and in doing so, to bring together for the first time the surprisingly large number of plasma researchers currently involved in work relating to coherent structures. The primary purpose of the workshop, in addition to the dissemination of information, was to develop formal and informal collaborations, set the stage for future formation of a coherent structures working group or focus area under the heading of the Tokamak Transport Task Force, and to evaluate the need for future workshops on coherent structures. The workshop was concentrated in four basic areas with a keynote talk in each area as well as 10 additional presentations. The issues of discussion in each of these areas was as follows: Theory - Develop a definition of structures and coherent as it applies to plasmas. Experiment - Review current experiments looking for structures in tokamaks, discuss experimental procedures for finding structures, discuss new experiments and techniques. Fluids - Determine how best to utilize the resource of information available from the fluids community both on the theoretical and experimental issues pertaining to coherent structures in plasmas. Computation - Discuss computational aspects of studying coherent structures in plasmas as they relate to both experimental detection and theoretical modeling

  13. Evolution of the spherical clusters

    International Nuclear Information System (INIS)

    Surdin, V.G.

    1978-01-01

    The possible processes of the Galaxy spherical clusters formation and evolution are described on a popular level. The orbits of spherical cluster motion and their spatial velocities are determined. Given are the distrbutions of spherical cluster stars according to their velocities and the observed distribution of spherical clusters in the area of the Galaxy slow evolution. The dissipation and dynamic friction processes destructing clusters with the mass less than 10 4 of solar mass and bringing about the reduction of clusters in the Galaxy are considered. The paradox of forming mainly X-ray sources in spherical clusters is explained. The schematic image of possible ways of forming X-ray sources in spherical clusters is given

  14. 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

  15. ELSA- The European Levitated Spherical Actruator

    Science.gov (United States)

    Ruiz, M.; Serin, J.; Telteu-Nedelcu, D.; De La Vallee Poussin, H.; Onillon, E.; Rossini, L.

    2014-08-01

    The reaction sphere is a magnetic bearing spherical actuator consisting of a permanent magnet spherical rotor that can be accelerated in any direction. It consists of an 8-pole permanent magnet spherical rotor that is magnetically levitated and can be accelerated about any axis by a 20-pole stator with electromagnets. The spherical actuator is proposed as a potential alternative to traditional momentum exchange devices such as reaction wheels (RWs) or control moment gyroscopes (CMGs). This new actuator provides several benefits such as reduced mass and power supply allocated to the attitude and navigation unit, performance gain, and improved reliability due to the absence of mechanical bearings. The paper presents the work done on the levitated spherical actuator and more precisely the electrical drive including its control unit and power parts. An elegant breadboard is currently being manufactured within the frame of an FP7 project. This project also comprises a feasibility study to show the feasibility of integrating such a system on a flight platform and to identify all the challenges to be solved in terms of technology or components to be developed.

  16. Measurement of Turbulence Modulation by Non-Spherical Particles

    DEFF Research Database (Denmark)

    Mandø, Matthias; Rosendahl, Lasse

    2010-01-01

    The change in the turbulence intensity of an air jet resulting from the addition of particles to the flow is measured using Laser Doppler Anemometry. Three distinct shapes are considered: the prolate spheroid, the disk and the sphere. Measurements of the carrier phase and particle phase velocities...... at the centerline of the jet are carried out for mass loadings of 0.5, 1, 1.6 and particle sizes 880μm, 1350μm, 1820μm for spherical particles. For each non-spherical shape only a single size and loading are considered. The turbulence modulation of the carrier phase is found to highly dependent on the turbulence......, the particle mass flow and the integral length scale of the flow. The expression developed on basis of spherical particles only is applied on the data for the non-spherical particles. The results suggest that non-spherical particles attenuate the carrier phase turbulence significantly more than spherical...

  17. Spherical grating spectrometers

    Science.gov (United States)

    O'Donoghue, Darragh; Clemens, J. Christopher

    2014-07-01

    We describe designs for spectrometers employing convex dispersers. The Offner spectrometer was the first such instrument; it has almost exclusively been employed on satellite platforms, and has had little impact on ground-based instruments. We have learned how to fabricate curved Volume Phase Holographic (VPH) gratings and, in contrast to the planar gratings of traditional spectrometers, describe how such devices can be used in optical/infrared spectrometers designed specifically for curved diffraction gratings. Volume Phase Holographic gratings are highly efficient compared to conventional surface relief gratings; they have become the disperser of choice in optical / NIR spectrometers. The advantage of spectrometers with curved VPH dispersers is the very small number of optical elements used (the simplest comprising a grating and a spherical mirror), as well as illumination of mirrors off axis, resulting in greater efficiency and reduction in size. We describe a "Half Offner" spectrometer, an even simpler version of the Offner spectrometer. We present an entirely novel design, the Spherical Transmission Grating Spectrometer (STGS), and discuss exemplary applications, including a design for a double-beam spectrometer without any requirement for a dichroic. This paradigm change in spectrometer design offers an alternative to all-refractive astronomical spectrometer designs, using expensive, fragile lens elements fabricated from CaF2 or even more exotic materials. The unobscured mirror layout avoids a major drawback of the previous generation of catadioptric spectrometer designs. We describe laboratory measurements of the efficiency and image quality of a curved VPH grating in a STGS design, demonstrating, simultaneously, efficiency comparable to planar VPH gratings along with good image quality. The stage is now set for construction of a prototype instrument with impressive performance.

  18. A spherical Taylor-Couette dynamo

    Science.gov (United States)

    Marcotte, Florence; Gissinger, Christophe

    2016-04-01

    We present a new scenario for magnetic field amplification in the planetary interiors where an electrically conducting fluid is confined in a differentially rotating, spherical shell (spherical Couette flow) with thin aspect-ratio. When the angular momentum sufficiently decreases outwards, a primary hydrodynamic instability is widely known to develop in the equatorial region, characterized by pairs of counter-rotating, axisymmetric toroidal vortices (Taylor vortices) similar to those observed in cylindrical Couette flow. We characterize the subcritical dynamo bifurcation due to this spherical Taylor-Couette flow and study its evolution as the flow successively breaks into wavy and turbulent Taylor vortices for increasing Reynolds number. We show that the critical magnetic Reynolds number seems to reach a constant value as the Reynolds number is gradually increased. The role of global rotation on the dynamo threshold and the implications for planetary interiors are finally discussed.

  19. Analysis of a spherical permanent magnet actuator

    International Nuclear Information System (INIS)

    Wang, J.; Jewell, G.W.; Howe, D.

    1997-01-01

    This paper describes a new form of actuator with a spherical permanent magnet rotor and a simple winding arrangement, which is capable of a high specific torque by utilizing a rare-earth permanent magnet. The magnetic-field distribution is established using an analytical technique formulated in spherical coordinates, and the results are validated by finite element analysis. The analytical field solution allows the prediction of the actuator torque and back emf in closed forms. In turn, these facilitate the characterization of the actuator and provide a firm basis for design optimization, system dynamic modeling, and closed-loop control law development. copyright 1997 American Institute of Physics

  20. 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.)