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Sample records for university hbt-ep tokamak

  1. Operational Performance of the Two-Channel 10 Megawatt Feedback Amplifier System for MHD Control on the Columbia University HBT-EP Tokamak

    International Nuclear Information System (INIS)

    Reass, W.A.; Wurden, G.A.

    1997-01-01

    The operational characteristics and performance of the two channel 10 Megawatt MHD feedback control system as installed by Los Alamos National Laboratory on the Columbia University HBT-EP tokamak are described. In the present configuration, driving independent 300 microH saddle coil sets, each channel can deliver 1100 Amperes and 16 kV peak to peak. Full power bandwidth is about 12 kHz, with capabilities at reduced power to 30 kHz. The present system topology is designed to suppress magnetohydrodynamic activity with m=2, n=1 symmetry. Application of either static (single phase) or rotating (twin phased) magnetic perturbations shows the ability to spin up or slow down the plasma, and also prevent (or cause) so-called ''mode-locking''. Open loop and active feedback experiments using a digital signal processor (DSP) have been performed on the HBT-EP tokamak and initial results show the ability to manipulate the plasma MHD mode frequency

  2. Design, construction, and first operational results of a 5 megawatt feedback controlled amplifier system for disruption control on the Columbia University HBT-EP tokamak

    International Nuclear Information System (INIS)

    Reass, W.A.; Alvestad, H.A.; Bartsch, R.R.; Wurden, G.A.; Ivers, T.H.; Nadle, D.L.

    1995-01-01

    This paper presents the electrical design and first operational results of a 5 Megawatt feedback controlled amplifier system designed to drive a 300 uH saddle coil set on the ''HBT-EP'' tokamak. It will be used to develop various plasma feedback techniques to control and inhibit the onset of plasma disruptions that are observed in high ''B'' plasmas. To provide a well characterized system, a high fidelity, high power closed loop amplifier system has been refurbished from the Los Alamos ''ZT-P'' equilibrium feedback system. In it's configuration developed for the Columbia HBT-EP tokamak, any desired waveform may be generated within a I 100 ampere and 16 kV peak to peak dynamic range. An energy storage capacitor bank presently limits the effective full power pulse width to 10 mS. The full power bandwidth driving the saddle coil set is ∼12 kHz, with bandwidth at reduced powers exceeding 30 kHz. The system is designed similar to a grounded cathode, push-pull, transformer coupled, tube type amplifier system. 'Me push pull amplifier consists of 6 each Machlett ML8618 magnetically beamed triodes, 3 on each end of the (center tapped) coupling transformer. The transformer has .I volt-seconds of core and a 1:1 turns ratio. The transformer is specially designed for high power, low leakage inductance, and high bandwidth. Each array of ML8618's is (grid) driven with a fiber optic controlled hotdeck with a 3CXI0,000A7 (triode) output. To linearize the ML8618 grid drive, a minor feedback loop in the hotdeck is utilized. Overall system response is controlled by active feedback of the saddle coil current, derived from a coaxial current viewing resistor. The detailed electrical design of the power amplifier, transformer, and feedback system will be provided in addition to recent HBT-EP operational results

  3. Passive stabilization of MHD instabilities at high βn in the HBT-EP Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gates, David A. [Columbia Univ., New York, NY (United States)

    1993-01-01

    The HBT-EP Tokamak has been designed, built, and is now fully operational in the Columbia University Plasma Physics Laboratory. One of the primary purposes of this facility is to study the effects of a conducting wall on the MHD modes that lead up to plasma disruptions. Of particular interest are the types of instabilities that are driven by the kinetic pressure of the plasma, because these instabilities are believed to be responsible for the present limit to plasma β with β ∝/B2, where the is the volume averaged pressure and B is the magnetic field. To this end, a movable conducting wall has been installed inside the HBT-EP vacuum chamber. The primary result of this thesis are the initial results from experiments that study the effect of this wall on plasma instabilities. The experiment shows that the conducting wall significantly reduces the growth rate of instabilities that precede a plasma disruption that occurs when the value of β is near the Troyon limit. The location of the wall required for significant stabilization is b/a ~1.2 where a is the minor radius of the plasma and b is the minor radial location of the wall. Moving the wall closer than b/a = 1.2 slightly degrades the stabilizing effect, which is consistent with recent theories.

  4. Passive stabilization of MHD instabilities at high βn in the HBT-EP Tokamak

    International Nuclear Information System (INIS)

    Gates, D.A.

    1993-01-01

    The HBT-EP Tokamak has been designed, built, and is now fully operational in the Columbia University Plasma Physics Laboratory. One of the primary purposes of this facility is to study the effects of a conducting wall on the MHD modes that lead up to plasma disruptions. Of particular interest are the types of instabilities that are driven by the kinetic pressure of the plasma, because these instabilities are believed to be responsible for the present limit to plasma β with β ∝ /B 2 , where the is the volume averaged pressure and B is the magnetic field. To this end, a movable conducting wall has been installed inside the HBT-EP vacuum chamber. The primary result of this thesis are the initial results from experiments that study the effect of this wall on plasma instabilities. The experiment shows that the conducting wall significantly reduces the growth rate of instabilities that precede a plasma disruption that occurs when the value of β is near the Troyon limit. The location of the wall required for significant stabilization is b/a ∼1.2 where a is the minor radius of the plasma and b is the minor radial location of the wall. Moving the wall closer than b/a = 1.2 slightly degrades the stabilizing effect, which is consistent with recent theories

  5. Active Control of 2/1 Magnetic Islands in the HBT-EP Tokamak

    International Nuclear Information System (INIS)

    Navratil, G.A.; Cates, C.; Mauel, M.E.; Maurer, D.; Nadle, D.; Taylor, E.; Xiao, Q.; Wurden, G.A.; Reass, W.A.

    1997-01-01

    Closed and open loop control techniques were applied to growing m/n=2/1 rotating islands in wall stabilized plasmas in the HBT-EP tokamak. The approach taken by HBT-EP combines an adjustable segmented conducting wall (which slows the growth or stabilizes ideal external kinks) with a number of small (6degree wide) saddle coils located between the gaps of the conducting wall. In this paper we report demonstration of 2-phase island rotation control from 5 kHz to 15 kHz and observation of the phase instability which are well modeled by the single-helicity, predictions of nonlinear Rutherford island dynamics for 2/1 tearing modes including important effects of ion inertia and FLR which appears as a damping term in the model equations. The closed loop response of active feedback control of the 2/1 mode at moderate gain was observed to be in good agreement with the theory. We have also demonstrated suppression of the 2/1 island growth using an asynchronous frequency modulation drive which maintains the flow damping of the island by application of rotating control fields with frequencies alternating above and below the natural mode frequency. This frequency modulation control technique was also able to prevent disruptions normally observed to follow giant sawtooth crashes in the plasma core

  6. Active feedback control of kink modes in tokamaks: 3D VALEN modeling and HBT-EP experiments

    International Nuclear Information System (INIS)

    Maurer, D.A.

    2002-01-01

    Significant progress in the development of active feedback control as a robust technique for the suppression of the wall stabilized external kink or resistive wall mode (RWM) in tokamaks has been achieved through a combination of modeling and experiments. Results from the 3D feedback modeling code VALEN, which serves as the primary analysis and feedback control design tool for RWM studies on the HBT-EP and DIII-D experiments, are in good agreement with observations. VALEN modeling of proposed advanced control system designs on HBT-EP, DIII-D, NSTX, and FIRE are predicted to approach the ideal wall beta limit in agreement with design principles based on simple single mode analytic theory of RWM feedback control. Benchmark experiments on HBT-EP have shown suppression of plasma disruption at rational edge q values using active feedback control in agreement with model predictions. In addition, the observation in HBT-EP of the plasma amplification of static resonant magnetic fields in plasmas marginally stable to the RWM is in agreement with theory. (author)

  7. High-speed, multi-input, multi-output control using GPU processing in the HBT-EP tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Rath, N., E-mail: Nikolaus@rath.org [Columbia University, Rm 200 Mudd, 500 W 120th St, New York, NY - 10027 (United States); Bialek, J.; Byrne, P.J.; DeBono, B.; Levesque, J.P.; Li, B.; Mauel, M.E.; Maurer, D.A.; Navratil, G.A.; Shiraki, D. [Columbia University, Rm 200 Mudd, 500 W 120th St, New York, NY - 10027 (United States)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer We present a GPU based system for magnetic control of perturbed equilibria. Black-Right-Pointing-Pointer Cycle times are below 5 {mu}s and I/O latencies below 10 {mu}s for 96 inputs and 64 outputs. Black-Right-Pointing-Pointer A new architecture removes host RAM and CPU from the control cycle. Black-Right-Pointing-Pointer GPU and DA/AD modules operate independently and communicate via PCIe peer-to-peer connections. Black-Right-Pointing-Pointer The Linux host system does not require real-time extensions. - Abstract: We report on the design of a new plasma control system for the HBT-EP tokamak that utilizes a graphical processing unit (GPU) to magnetically control the 3D perturbed equilibrium state [1] of the plasma. The control system achieves cycle times of 5 {mu}s and I/O latencies below 10 {mu}s for up to 96 inputs and 64 outputs. The number of state variables is in the same order. To handle the resulting computational complexity under the given time constraints, the control algorithms are designed for massively parallel processing. The necessary hardware resources are provided by an NVIDIA Tesla M2050 GPU, offering a total of 448 computing cores running at 1.3 GHz each. A new control architecture allows control input from magnetic diagnostics to be pushed directly into GPU memory by a D-TACQ ACQ196 digitizer, and control output to be pulled directly from GPU memory by two D-TACQ AO32 analog output modules. By using peer-to-peer PCI express connections, this technique completely eliminates the use of host RAM and central processing unit (CPU) from the control cycle, permitting single-digit microsecond latencies on a standard Linux host system without any real-time extensions.

  8. Initial Ferritic Wall Mode studies on HBT-EP

    Science.gov (United States)

    Hughes, Paul; Bialek, J.; Boozer, A.; Mauel, M. E.; Levesque, J. P.; Navratil, G. A.

    2013-10-01

    Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective US component test facility and DEMO. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these experiments. Although the ferritic wall mode (FWM) was seen in a linear machine, the FWM was not observed in JFT-2M, probably due to eddy current stabilization. Using its high-resolution magnetic diagnostics and positionable walls, HBT-EP has begun exploring the dynamics and stability of plasma interacting with high-permeability ferritic materials tiled to reduce eddy currents. We summarize a simple model for plasma-wall interaction in the presence of ferromagnetic material, describe the design of a recently-installed set of ferritic shell segments, and report initial results. Supported by U.S. DOE Grant DE-FG02-86ER53222.

  9. [High beta tokamak research and plasma theory

    International Nuclear Information System (INIS)

    1990-01-01

    Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report

  10. High-Speed imaging of the plasma response to resonant magnetic perturbations in HBT-EP

    International Nuclear Information System (INIS)

    Angelini, Sarah M; Levesque, Jeffrey P; Mauel, Michael E; Navratil, Gerald A

    2015-01-01

    A Phantom v7.3 fast digital camera was used to study visible light fluctuations in the High Beta Tokamak–Extended Pulse (HBT–EP). This video data is the first to be used to analyze and understand the behavior of long wavelength kink perturbations in a wall-stabilized tokamak. The light was mostly comprised of Dα 656 nm light. Profiles of the plasma light at the midplane were hollow with a radial scale length of approximately 4 cm at the plasma edge. The fast camera was also used to measure the plasma’s response to applied helical magnetic perturbations. The programmed toroidal phase angle of the resonant magnetic perturbation (RMP) was directly inferred from the resulting images of the plasma response. The plasma response and the intensity of the RMP were compared under different conditions. The resulting amplitude correlations are consistent with previous measurements of the static response using an array of magnetic sensors. (paper)

  11. The influence of a conducting wall on disruptions in HBT-EP

    International Nuclear Information System (INIS)

    Kombargi, R.

    1997-01-01

    The characteristics of long wavelength magnetohydrodynamic (MHD) disruptive instabilities have been studied in a tokamak device with segmented and movable conducting walls. Coupling between the wall and the plasma was varied by systematically adjusting the radial position, b, of the conducting plates relative to the plasma surface of minor radius a. By pre-selecting the total plasma current ramp rate (dI p /dt), disruptive instabilities driven either by large edge currents or high plasma pressure were studied. Specifically, three types of disruptions caused by both external (to the plasma) and internal instabilities were obtained by changing the plasma-wall separation (b/a) and the temporal evolution of the total plasma current. The properties of these disruptions were examined using a variety of magnetic pickup coils and arrays of soft x-ray detectors. Experiments demonstrated that rapidly developing, low-n kink instabilities were suppressed if the conducting wall was positioned sufficiently near the plasma (b/a p /dt > 0) was maintained. Conducting wall stabilization of fast growing external instabilities was observed in discharges with high edge current and in plasmas with β-values near the ideal MED stability boundary. When the conducting wall was near the plasma surface and as the current profile evolved in time (dI p /dt < 0), slowly growing internal instabilities would also lead to disruptions. Disruption mechanisms for plasmas with b/a=1.52 and b/a=1.07 were compared. Differences in the precursor modes, the speed of the thermal collapse and the chronological sequence of events were found. In summary, the disruptions with an external character were eliminated when the conducting wall was moved from b/a=1.52 to b/a<1.2. Internal disruptions could not be averted even with b/a=1.07

  12. The high beta tokamak-extended pulse magnetohydrodynamic mode control research program

    International Nuclear Information System (INIS)

    Maurer, D A; Bialek, J; Byrne, P J; De Bono, B; Levesque, J P; Li, B Q; Mauel, M E; Navratil, G A; Pedersen, T S; Rath, N; Shiraki, D

    2011-01-01

    The high beta tokamak-extended pulse (HBT-EP) magnetohydrodynamic (MHD) mode control research program is studying ITER relevant internal modular feedback control coil configurations and their impact on kink mode rigidity, advanced digital control algorithms and the effects of plasma rotation and three-dimensional magnetic fields on MHD mode stability. A new segmented adjustable conducting wall has been installed on the HBT-EP and is made up of 20 independent, movable, wall shell segments instrumented with three distinct sets of 40 saddle coils, totaling 120 in-vessel modular feedback control coils. Each internal coil set has been designed with varying toroidal angular coil coverage of 5, 10 and 15 0 , spanning the toroidal angle range of an ITER port plug based internal coil to test resistive wall mode (RWM) interaction and multimode MHD plasma response to such highly localized control fields. In addition, we have implemented 336 new poloidal and radial magnetic sensors to quantify the applied three-dimensional fields of our control coils along with the observed plasma response. This paper describes the design and implementation of the new control shell incorporating these control and sensor coils on the HBT-EP, and the research program plan on the upgraded HBT-EP to understand how best to optimize the use of modular feedback coils to control instability growth near the ideal wall stabilization limit, answer critical questions about the role of plasma rotation in active control of the RWM and the ferritic resistive wall mode, and to improve the performance of MHD control systems used in fusion experiments and future burning plasma systems.

  13. 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

  14. 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

  15. 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

  16. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-05-01

    discharges, production and self-organization of a turbulent plasma column in a spheromak (''SK-CG-1''), and (iv) a planned large-aspect ratio, high-beta tokamak (HBT-EP) experiment. Refs, figs and tabs

  17. High resolution detection and excitation of resonant magnetic perturbations in a wall-stabilized tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Maurer, David A. [Physics Department, Auburn University, Auburn, Alabama 36849 (United States); Shiraki, Daisuke; Levesque, Jeffrey P.; Bialek, James; Angelini, Sarah; Byrne, Patrick; DeBono, Bryan; Hughes, Paul; Mauel, Michael E.; Navratil, Gerald A.; Peng Qian; Rhodes, Dov; Rath, Nickolaus; Stoafer, Christopher [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2012-05-15

    We report high-resolution detection of the 3D plasma magnetic response of wall-stabilized tokamak discharges in the High Beta Tokamak-Extended Pulse [T. H. Ivers et al., Phys. Plasmas 3, 1926 (1996)] device. A new adjustable conducting wall has been installed on HBT-EP made up of 20 independent, movable, wall segments instrumented with three distinct sets of 40 modular coils that can be independently driven to generate a wide variety of magnetic perturbations. High-resolution detection of the plasma response is made with 216 poloidal and radial magnetic sensors that have been located and calibrated with high-accuracy. Static and dynamic plasma responses to resonant and non-resonant magnetic perturbations are observed through measurement of the step-response following a rapid change in the toroidal phase of the applied perturbations. Biorthogonal decomposition of the full set of magnetic sensors clearly defines the structures of naturally occurring external kinks as being composed of independent m/n = 3/1 and 6/2 modes. Resonant magnetic perturbations were applied to discharges with pre-existing, saturated m/n = 3/1 external kink mode activity. This m/n = 3/1 kink mode was observed to lock to the applied perturbation field. During this kink mode locked period, the plasma resonant response is characterized by a linear, a saturated, and a disruptive plasma regime dependent on the magnitude of the applied field and value of the edge safety factor and plasma rotation.

  18. MHD Effects of a Ferritic Wall on Tokamak Plasmas

    Science.gov (United States)

    Hughes, Paul E.

    It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency

  19. On the universality of power laws for tokamak plasma predictions

    Science.gov (United States)

    Garcia, J.; Cambon, D.; Contributors, JET

    2018-02-01

    Significant deviations from well established power laws for the thermal energy confinement time, obtained from extensive databases analysis as the IPB98(y,2), have been recently reported in dedicated power scans. In order to illuminate the adequacy, validity and universality of power laws as tools for predicting plasma performance, a simplified analysis has been carried out in the framework of a minimal modeling for heat transport which is, however, able to account for the interplay between turbulence and collinear effects with the input power known to play a role in experiments with significant deviations from such power laws. Whereas at low powers, the usual scaling laws are recovered with little influence of other plasma parameters, resulting in a robust power low exponent, at high power it is shown how the exponents obtained are extremely sensitive to the heating deposition, the q-profile or even the sampling or the number of points considered due to highly non-linear behavior of the heat transport. In particular circumstances, even a minimum of the thermal energy confinement time with the input power can be obtained, which means that the approach of the energy confinement time as a power law might be intrinsically invalid. Therefore plasma predictions with a power law approximation with a constant exponent obtained from a regression of a broad range of powers and other plasma parameters which can non-linearly affect and suppress heat transport, can lead to misleading results suggesting that this approach should be taken cautiously and its results continuously compared with modeling which can properly capture the underline physics, as gyrokinetic simulations.

  20. A universal access layer for the Integrated Tokamak Modelling Task Force

    International Nuclear Information System (INIS)

    Manduchi, G.; Iannone, F.; Imbeaux, F.; Huysmans, G.; Lister, J.B.; Guillerminet, B.; Strand, P.; Eriksson, L.-G.; Romanelli, M.

    2008-01-01

    The Integrated Tokamak Modelling (ITM) Task Force aims at providing a suite of codes for preparing and analyzing future ITER discharges. In the framework of the ITM, the universal access layer (UAL) provides the capability of storing and retrieving data involved in simulation. The underlying data structure is hierarchical and the granularity in data access is given by the definition of a set of consistent physical objects (CPOs). To describe the data structure of the overall ITM database, the XML schema description (XSD) has been used. Originally intended to describe the structure of XML documents, XSD is used here to provide an unambiguous way of describing how data are structured, regardless of the actual implementation of the underlying database. The MDSplus-based UAL implementation is currently under test and other prototypes for investigating alternative data storage systems are foreseen

  1. 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

  2. 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

  3. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

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

    2012-04-01

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

  4. 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

  5. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

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

  6. 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

  7. 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

  8. 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)

  9. Plasma position control in TCABR Tokamak

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  10. Design and installation of a ferromagnetic wall in tokamak geometry

    International Nuclear Information System (INIS)

    Hughes, P. E.; Levesque, J. P.; Rivera, N.; Mauel, M. E.; Navratil, G. A.

    2015-01-01

    Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak–Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics and overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability

  11. Design and installation of a ferromagnetic wall in tokamak geometry

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P. E., E-mail: peh2109@columbia.edu; Levesque, J. P.; Rivera, N.; Mauel, M. E.; Navratil, G. A. [Columbia University Plasma Physics Laboratory, Columbia University, 102 S.W. Mudd, 500 W. 120th St., New York, New York 10027 (United States)

    2015-10-15

    Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak–Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics and overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability.

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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.

  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. 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

  2. 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

  3. Heat load material studies: Simulated tokamak disruptions

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  4. 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)

  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. 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

  7. 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)

  8. 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.

  9. 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

  10. Multi-mode remote participation on the GOLEM tokamak

    International Nuclear Information System (INIS)

    Svoboda, V.; Huang, B.; Mlynar, J.; Pokol, G.I.; Stoeckel, J.; Vondrasek, G.

    2011-01-01

    The GOLEM tokamak (formerly CASTOR) at Czech Technical University is demonstrated as an educational tokamak device for domestic and foreign students. Remote participation of several foreign universities (in Hungary, Belgium, Poland and Costa Rica) has been successfully performed. A unique feature of the GOLEM device is functionality which enables complete remote participation and control, solely through Internet access. Basic remote control is possible either in online mode via WWW/SSH interface or offline mode using batch processing code. Discharge parameters are set in each case to configure the tokamak for a plasma discharge. Using the X11 protocol it is possible to control in an advanced mode many technological aspects of the tokamak operation, including: i) vacuum pump initialization, ii) chamber baking, iii) charging of power supplies, iv) plasma discharge scenario, v) data acquisition system.

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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.

  14. Runaway-ripple interaction in Tokamaks

    International Nuclear Information System (INIS)

    Laurent, L.; Rax, J.M.

    1989-08-01

    Two approaches of the interaction between runaway electrons and the ripple field, in tokamaks, are discussed. The first approach considers the resonance effect as an intense cyclotron heating of the electrons, by the ripple field, in the guiding center frame of the fast particles. In the second approach, an Hamiltonian formalism is used. A criterion for the onset of chaotic behavior and the results are given. A new universal instability of the runaway population in tokamak configuration is found. When combined with cyclotron losses one of its major consequence is to act as an effective slowing down mechanism preventing the free fall acceleration toward the synchrotron limit. This configuration allows the explanation of some experimental results of Tore Supra and Textor

  15. 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

  16. 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

  17. 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.)

  18. 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.

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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.)

  4. 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

  5. Experimental results from the TUMAN 3 tokamak

    International Nuclear Information System (INIS)

    Golant, V.E.; Andrejko, M.V.; Askinazi, L.G.; Korneev, V.A.; Krikunov, S.V.; Lipin, B.M.; Lebedev, S.V.; Levin, L.S.; Podushnikova, K.A.; Razdobarin, G.T.; Rozhansky, V.A.; Rozhdestvensky, V.V.; Tendler, M.; Tukachinsky, A.S.; Jaroshevich, S.P.

    1995-01-01

    The open-quote open-quote TUMAN-3 close-quote close-quote Tokamak programme concentrates on issues of improved confinement. In 1989 the transition from an ordinary Ohmic regime into an improved confinement mode was achieved. The signatures of the H-mode in auxiliary heated tokamaks have been observed in this regime. The crucial role of the boundary radial electric field was found in the experiments with internal bias probe. Other techniques were demonstrated to disturb the boundary plasma which led to H-mode triggering: short increase of working gas puffing, minor radius magnetic compression and pellet injection. The role scaling of the energy confinement time in the Ohmic H-mode was obtained, which differs dramatically from the scaling for the ordinary Ohmic regime. There were found a strong dependence of τ E on plasma current and a weak dependence on density. The maximum value of τ E was 10 times longer than in the ordinary Ohmic region. The τ E scaling for the Ohmic H-mode is consistent with the scaling proposed for devices with powerful auxiliary heating. The results shows that H-mode physics is universal in tokamaks with different geometries and heating methods. (AIP) copyright 1995 American Institute of Physics

  6. 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.

  7. 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

  8. 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

  9. MHD stability of an almost circular tokamak

    International Nuclear Information System (INIS)

    Roy, A.

    1990-10-01

    In a tokamak, the ratio β between the plasma pressure and that of the magnetic field is limited by the appearance of instabilities. The magnetic field in a tokamak reactor will always be limited by technological constraints. It is therefore crucial to know what factors have an effect on the β limit, since a zero resistivity plasma fluid model allows for theoretical reproduction of the β limits observed experimentally. Theoretical studies have shown that the distributions of pressure and current density may have a substantial effect on the β limit. The effect of the current density and pressure distributions on the β limit has been studied for tokamak with a circular core section. The best results are obtained when the current density is concentrated in the centre of the section and is nil at the periphery. But the second region of stability against ballooning modes cannot be obtained in a circular tokamak owing to the destabilisation of the universal modes. This study was then extended to the stability of plasmas the section of which is almost circular and has a point of reflection. Such configurations are vital for fusion since they allow systems in which the confinement time does not deteriorate with an increase in the additional heating power. The β limit was calculated for different positions of the reflection point. The results show that when it is displaced from the interior towards the exterior of the torus, the stability of the overall modes is progressively improved until it is vertical. But if the point of reflection is further displaced from this vertical position towards the exterior of the torus, localised modes close to the edge of the plasma are destabilised and bring about a drop in the β limit. (author) figs., tabs., 80 refs

  10. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. TBR-1 (Brazilian Tokamak) - Recent Results

    International Nuclear Information System (INIS)

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

    1987-08-01

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

  16. Present status of TCA/BR Tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Galvao, R.M.O.; Tuszel, A.G.

    1997-01-01

    The TCA tokamak is being partially reconstructed and reassembled in the Plasma Laboratory of The University of Sao Paulo, and afterwards it will be named TCA/BR. The first discharges are expected by June/July of next year. The main scientific objectives envisaged for the machine are: Alfven wave heating and current drive, confinement improvement, disruptions and turbulence. In this paper we also describe: (i) the present status of the project; (ii) the diagnostic system; (iii) the control and data acquisition system; (iv) the RF system for the excitation of Alfven waves, that are being developed, and also the results of predictive transport simulations of its performance. (author)

  17. 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

  18. 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

  19. 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

  20. 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

  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. 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)

  3. 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

  4. 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.)

  5. Alcator C-Mod Tokamak

    Data.gov (United States)

    Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

  6. 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)

  7. New directions in tokamak reactors

    International Nuclear Information System (INIS)

    Baker, C.C.

    1985-01-01

    New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs

  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. Numerical Tokamak Project code comparison

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  10. 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)

  11. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

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

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. Anomalous transport in tokamaks

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1989-01-01

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

  17. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-09-01

    A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  18. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-01-01

    A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  19. 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

  20. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)

  1. Disruptions in Tokamaks

    International Nuclear Information System (INIS)

    Bondeson, A.

    1987-01-01

    This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)

  2. 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

  3. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Hively, L.M.; Sigmar, D.J.

    1989-09-01

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  4. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  5. 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)

  6. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

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

  7. 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

  8. 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.

  9. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

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

  10. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

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

    2012-01-01

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  11. Resistive instabilities in tokamaks

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1985-10-01

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

  12. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  13. 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)

  14. 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

  15. Compact Ignition Tokamak conventional facilities optimization

    International Nuclear Information System (INIS)

    Commander, J.C.; Spang, N.W.

    1987-01-01

    A high-field ignition machine with liquid-nitrogen-cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed for the next phase of the United States magnetically confined fusion program. A team of national laboratory, university, and industrial participants completed the conceptual design for the CIT machine, support systems and conventional facilities. Following conceptual design, optimization studies were conducted with the goal of improving machine performance, support systems design, and conventional facilities configuration. This paper deals primarily with the conceptual design configuration of the CIT conventional facilities, the changes that evolved during optimization studies, and the revised changes resulting from functional and operational requirements (F and ORs). The CIT conventional facilities conceptual design is based on two premises: (1) satisfaction of the F and ORs developed in the CIT building and utilities requirements document, and (2) the assumption that the CIT project will be sited at the Princeton Plasma Physics Laboratory (PPPL) in order that maximum utilization can be made of existing Tokamak Fusion Test Reactor (TFTR) buildings and utilities. The optimization studies required reevaluation of the F and ORs and a second look at TFTR buildings and utilities. Some of the high-cost-impact optimization studies are discussed, including the evaluation criteria for a change from the conceptual design baseline configuration. The revised conventional facilities configuration are described and the estimated cost impact is summarized

  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. Universe

    CERN Document Server

    2009-01-01

    The Universe, is one book in the Britannica Illustrated Science Library Series that is correlated to the science curriculum in grades 5-8. The Britannica Illustrated Science Library is a visually compelling set that covers earth science, life science, and physical science in 16 volumes.  Created for ages 10 and up, each volume provides an overview on a subject and thoroughly explains it through detailed and powerful graphics-more than 1,000 per volume-that turn complex subjects into information that students can grasp.  Each volume contains a glossary with full definitions for vocabulary help and an index.

  18. Conceptual Design of Alborz Tokamak Poloidal Coils System

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.

    2013-04-01

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

  19. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  20. Large aspect ratio tokamak study

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

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

  2. 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)

  3. 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

  4. 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

  5. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.

    1990-10-01

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  6. 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.)

  7. Design approach of seismic interface for cryoline with Tokamak building for ITER

    International Nuclear Information System (INIS)

    Badgujar, S.; Sarkar, B.; Vaghela, H.; Shah, N.; Naik, H.B.

    2012-01-01

    ITER Tokamak building is designed with seismic isolation pads to protect the Tokamak components from seismic events. Two main cryolines, designated as cryolines between buildings (Mg and CP), runs from interconnection box in cryoplant building to the Tokamak building. The lines outside Tokamak building are supported by seismically non-isolated supports. The cryoline design at the interface between seismically isolated and non-isolated support systems needs to be studied to fulfill the functional requirements. One of the options for interface, universal expansion joint has been modeled in CATIA with actual thickness of each ply and inter-ply distance, analyzed in ANSYS using contact definition, as a part of the preliminary study. The bellows have been checked by design calculation as per EJMA standard for the specified movements. The paper will present approach for conceptual design of interface, problem definition and boundary conditions, methodology for analysis and preliminary results of stress pattern for expansion joints. (author)

  8. 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 ...

  9. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

    Mantsinen, M.J.

    1995-03-01

    The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)

  10. Energy losses on tokamak startup

    International Nuclear Information System (INIS)

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1983-01-01

    During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells

  11. 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

  12. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2011-01-01

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

  13. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

  14. Tokamak and RFP ignition requirements

    International Nuclear Information System (INIS)

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

  15. Tokamak impurity-control techniques

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1980-01-01

    A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)

  16. 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.

  17. Multimegawatt neutral beams for tokamaks

    International Nuclear Information System (INIS)

    Kunkel, W.B.

    1979-03-01

    Most of the large magnetic confinement experiments today and in the near future use high-power neutral-beam injectors to heat the plasma. This review briefly describes this remarkable technique and summarizes recent results as well as near term expectations. Progress has been so encouraging that it seems probable that tokamaks will achieve scientific breakeven before 1990

  18. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  19. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-01-01

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

  20. 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)

  1. Heating of plasmas in tokamaks by current-driven turbulence

    International Nuclear Information System (INIS)

    Kluiver, H. de.

    1985-10-01

    Investigations of current-driven turbulence have shown the potential to heat plasmas to elevated temperatures in relatively small cross-section devices. The fundamental processes are rather well understood theoretically. Even as it is shown to be possible to relax the technical requirements on the necessary electric field and the pulse length to acceptable values, the effect of energy generation near the plasma edge, the energy transport, the impurity influx and the variation of the current profile are still unknown for present-day large-radius tokamaks. Heating of plasmas by quasi-stationary weakly turbulent states caused by moderate increases of the resistivity due to higher loop voltages could be envisaged. Power supplies able to furnish power levels 5-10 times higher than the usual values could be used for a demonstration of those regimes. At several institutes and university laboratories the study of turbulent heating in larger tokamaks and stellarators is pursued

  2. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

    1990-05-01

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

  3. Combined confinement system applied to tokamaks

    International Nuclear Information System (INIS)

    Ohkawa, Tihiro

    1986-01-01

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

  4. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  5. Improvement of the tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, L

    1994-12-31

    Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.

  6. 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

  7. Flux driven turbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.

    1999-01-01

    This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)

  8. Options for an ignited tokamak

    International Nuclear Information System (INIS)

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon β/sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed

  9. 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

  10. 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

  11. 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

  12. Magnetic island formation in tokamaks

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs

  13. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

    Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)

  14. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

    The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

  15. Runaway electrons during tokamak startup

    International Nuclear Information System (INIS)

    Sharma, A.S.; Jayakumar, R.

    1988-01-01

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

  16. Minimum scaling laws in tokamaks

    International Nuclear Information System (INIS)

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

    1986-10-01

    Scaling laws governing anomalous electron transport in tokamaks with ohmic and/or auxiliary heating are derived using renormalized Vlasov-Ampere equations for low frequency electromagnetic microturbulence. It is also shown that for pure auxiliary heating (or when auxiliary heating power far exceeds the ohmic power), the energy confinement time scales as tau/sub E/ ∼ P/sub inj//sup -1/3/, where P/sub inj/ is the injected power

  17. Gyrosheath near the tokamak edge

    International Nuclear Information System (INIS)

    Hazeltine, R.D.; Xiao, H.; Valanju, P.M.

    1993-03-01

    A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results

  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. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Kieras, C.E.

    1982-12-01

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

  20. Discharge cleaning for a tokamak

    International Nuclear Information System (INIS)

    Ishii, Shigeyuki

    1983-01-01

    Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)

  1. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

  2. Plasma edge physics in an actively cooled tokamak

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  3. Convective transport in tokamaks

    International Nuclear Information System (INIS)

    D'Ippolito, D.A.; Myra, J.R.; Russell, D.A.; Krasheninnikov, S.I.; Pigarov, A.Yu.; Yu, G.Q.; Xu, X.Q.; Nevins, W.M.

    2005-01-01

    Scrape-off-layer (SOL) convection in fusion experiments appears to be a universal phenomenon that can 'short-circuit' the divertor in some cases. The theory of 'blob' transport provides a simple and robust physical paradigm for studying convective transport. This paper summarizes recent advances in the theory of blob transport and its comparison with 2D and 3D computer simulations. We also discuss the common physical basis relating radial transport of blobs, pellets, and ELMs and a new blob regime that may lead to a connection between blob transport and the density limit. (author)

  4. Report on the high magnetic field tokamak TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-02-01

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

  5. 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.)

  6. Magnetic confinement by Tokamak: physical aspects

    International Nuclear Information System (INIS)

    Tachon, J.

    1980-01-01

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

  7. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    Reid, R.L.; Steiner, D.

    1977-01-01

    This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  8. Simulation of a major tokamak disruption

    International Nuclear Information System (INIS)

    White, R.B.; Monticello, D.A.; Rosenbluth, M.N.

    1977-08-01

    It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption

  9. Diagnostics for the Rijnhuizen Tokamak Project

    NARCIS (Netherlands)

    Donne, A. J. H.

    1994-01-01

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

  10. 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.)

  11. Mercier criterion for high-β tokamaks

    International Nuclear Information System (INIS)

    Galvao, R.M.O.

    1984-01-01

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

  12. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  13. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  14. Design, simulation and construction of the Taban tokamak

    Science.gov (United States)

    H, R. MIRZAEI; R, AMROLLAHI

    2018-04-01

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

  15. Turbulent transport reduction by E x B velocity shear during edge plasma biasing in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Van Oost, G. [Dept. of Applied Physics, Ghent Univ., Ghent (Belgium); Adamek, J.; Antoni, V.; Balan, P.; Boedo, J.A.; Devynck, P.; Duran, I.; Eliseev, L.; Gunn, J.P.; Hron, M.; Ionita, C.; Jachmich, S.; Kirnev, G.S.; Martines, E.; Melnikov, A.; Peleman, P.; Schrittwieser, R.; Silva, C.; Stoeckel, J.; Tendler, M.; Varandas, C.; Van Schoor, M.; Vershkov, V.; Weynants, R.R.

    2004-07-01

    Experiments in the tokamaks TEXTOR, CASTOR, T-10 and ISTTOK have provided new and complementary evidence on the physics of the universal mechanism of E x B velocity shear stabilization of turbulence, concomitant transport barrier formation and radial conductivity by using various edge biasing techniques. (orig.)

  16. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír

    2012-01-01

    Roč. 87, č. 12 (2012), s. 2238-2241 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006

  17. Surface tearing modes in tokamaks

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Kurita, Gen-ichi; Azumi, Masafumi; Takeda, Tatsuoki

    1985-10-01

    Surface tearing modes in tokamaks are studied numerically and analytically. The eigenvalue problem is solved to obtain the growth rate and the mode structure. We investigate in detail dependences of the growth rate of the m/n = 2/1 resistive MHD modes on the safety factor at the plasma surface, current profile, wall position, and resistivity. The surface tearing mode moves the plasma surface even when the wall is close to the surface. The stability diagram for these modes is presented. (author)

  18. Major disruption process in tokamak

    International Nuclear Information System (INIS)

    Kurita, Gen-ichi; Azumi, Masafumi; Tuda, Takashi; Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji; Itoh, Kimitaka; Takeda, Tatsuoki

    1981-11-01

    The major disruption in a cylindrical tokamak is investigated by using the multi-helicity code, and the destabilization of the 3/2 mode by the mode coupling with the 2/1 mode is confirmed. The evolution of the magnetic field topology caused by the major disruption is studied in detail. The effect of the internal disruption on the 2/1 magnetic island width is also studied. The 2/1 magnetic island is not enhanced by the flattening of the q-profile due to the internal disruption. (author)

  19. Final technical report for DE-SC00012633 AToM (Advanced Tokamak Modeling)

    Energy Technology Data Exchange (ETDEWEB)

    Holland, Christopher [Univ. of California, San Diego, CA (United States); Orlov, Dmitri [Univ. of California, San Diego, CA (United States); Izzo, Valerie [Univ. of California, San Diego, CA (United States)

    2018-02-05

    This final report for the AToM project documents contributions from University of California, San Diego researchers over the period of 9/1/2014 – 8/31/2017. The primary focus of these efforts was on performing validation studies of core tokamak transport models using the OMFIT framework, including development of OMFIT workflow scripts. Additional work was performed to develop tools for use of the nonlinear magnetohydrodynamics code NIMROD in OMFIT, and its use in the study of runaway electron dynamics in tokamak disruptions.

  20. Computer and engineering calculations of Brazilian Tokamak-II

    International Nuclear Information System (INIS)

    Wang, S.; Chen, Y.; Sa, W.P. de; Nascimento, I.C.; Tuszel, A.G.; Galvao, R.M.O.; Machida, M.

    1990-01-01

    Analytical and computer calculations carried out by researches of Physics Institute - University of Sao Paulo (IFUSP), for defining the engineering project and constructing the TBR-II tokamak are presented. The hydrodynamics behavioue and determined parameters for magnetic confinement of the plasma were analysed. The computer code was developed using magnetohydrodynamics (MHD) equations which involve plasma interactions, magnetic field and electrical current circulating in more than 20 coils distributed around toroidal vase of the plasma. The electromagnetic, thermal and mechanical couplings are also presented. The TBR-II will be feed by two turbo-generators with 15 MW each one. (M.C.K.) [pt

  1. Simulation experiment on magnetic field reconnection processes in tokamak

    International Nuclear Information System (INIS)

    Kiwamoto, Y.

    1982-01-01

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

  2. Minerals resource implications of a tokamak fusion reactor economy

    Energy Technology Data Exchange (ETDEWEB)

    Cameron, E; Conn, R W; Kulcinski, G L; Sviatoslavsky, I

    1979-09-01

    The mineral resource implications of an economy of tokamak-type fusion reactors are assessed based upon the recent conceptual reactor design study, NUWMAK, developed at the University of Wisconsin. For comparative purposes, various structural alloys of vanadium and steel are assumed to be usable in the NUWMAK design in place of the titanium alloy originally selected. In addition, the inner blanket core and magnet system of the conceptual reactor, HFCTR, developed at the Massachusetts Institute of Technology, are assumed to be interchangeable with the comparable components in NUWMAK. These variations permit a range of likely requirements to be assessed.

  3. Minerals resource implications of a tokamak fusion reactor economy

    International Nuclear Information System (INIS)

    Cameron, E.; Conn, R.W.; Kulcinski, G.L.; Sviatoslavsky, I.

    1979-09-01

    The mineral resource implications of an economy of tokamak-type fusion reactors are assessed based upon the recent conceptual reactor design study, NUWMAK, developed at the University of Wisconsin. For comparative purposes, various structural alloys of vanadium and steel are assumed to be usable in the NUWMAK design in place of the titanium alloy originally selected. In addition, the inner blanket core and magnet system of the conceptual reactor, HFCTR, developed at the Massachusetts Institute of Technology, are assumed to be interchangeable with the comparable components in NUWMAK. These variations permit a range of likely requirements to be assessed

  4. Conceptual design of HL-2M tokamak control system

    International Nuclear Information System (INIS)

    Xia Fan; Chen Liaoyuan; Song Xianming; Zhang Jinhua; Lou Cuiwen; Pan Yudong

    2009-01-01

    The static architecture, dynamic behavior, control theory and simulation of HL-2M tokamak control system are described. The real-time network will be build for the communication of real-time control among its subsystems and universal timing system will be build to guarantee the synchronization among the subsystems. The duty to achieve preprogrammed parameters is carried out by plasma discharge control. In order to reduce the damage made by discharge exception, the error-handing mechanism of supervision system is considered. The controllers of magnetic control system are designed to control the current, shape and position of plasma and simulation system is designed for testing the controllers. (authors)

  5. Vertical injection of compact torus into the STOR-M tokamak

    International Nuclear Information System (INIS)

    Liu, D.; Singh, A.K.; Hirose, A.; Xiao, C.

    2005-01-01

    Vertical compact torus injection into the STOR-M tokamak has been conducted with the University of Saskatchewan Compact Torus Injector (USCTI). The injector stayed at the horizontal position and the CT was bent by 90 deg. using a curved conducting drift tube. The curved drift tube did not have significant effects on the CT velocity. Furthermore, the curved drift tube did not change the magnetic field topology. Preliminary vertical CT injection experiments have been carried out on the STOR-M tokamak. CT injection induced prompt increase in the electron density and in the soft x-ray radiation level. Further modifications of the 90 deg. are underway to improve the CT parameters and to further study the effects of CT injection on the tokamak plasma parameters. (author)

  6. 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.)

  7. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Energy Technology Data Exchange (ETDEWEB)

    Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Kocman, J.; Grover, O. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Krbec, J.; Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, CZ-182 21 Prague (Czech Republic)

    2015-10-15

    Graphical abstract: * Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes.* Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform.* More than 20% plasma life prolongation with plasma position control in feedback mode. - Highlights: • Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes. • Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform. • More than 20% plasma life prolongation with plasma position control in feedback mode. - Abstract: The GOLEM tokamak at the Czech Technical University has been established as an educational tokamak device for domestic and foreign students. Remote participation in the scope of several laboratory practices, plasma physics schools and workshops has been successfully performed from abroad. A new enhancement allowing understandable remote control of vertical plasma position in two modes (i) predefined and (ii) feedback control is presented. It allows to drive the current in the stabilization coils in any time-dependent scenario, which can include as a parameter the actual plasma position measured by magnetic diagnostics. Arbitrary movement of the plasma column in a vertical direction, stabilization of the plasma column in the center of the tokamak vessel as well as prolongation/shortening of plasma life according to the remotely defined request are demonstrated.

  8. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-05-01

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime

  9. CAT-D-T tokamaks

    International Nuclear Information System (INIS)

    Greenspan, E.; Blue, T.; Miley, G.H.

    1981-01-01

    The domains of plasma fuel cycles bounded by the D-T and Cat-D, and by the D-T and SCD modes of operation are examined. These domains, referred to as, respectively, the Cat-D-T and SCD-T modes of operation, are characterized by the number (γ) of tritons per fusion neutron available from external (to the plasma) sources. Two external tritium sources are considered - the blankets of the Cat-D-T (SCD-T) reactors and fission reactors supported by the Cat-D-T (SCD-T) driven hybrid reactors. It is found that by using 6 Li for the active material of the control elements of the fission reactors, it is possible to achieve γ values close to unity. Cat-D-T tokamaks could be designed to have smaller size, higher power density, lower magnetic field and even lower plasma temperature than Cat-D tokamaks; the difference becomes significant for γ greater than or equal to .75. The SCD-T mode of operation appears to be even more attractive. Promising applications identified for these Cat-D-T and SCD-T modes of operation include hybrid reactors, fusion synfuel factories and fusion reactors which have difficulty in providing all their tritium needs

  10. How much does a tokamak reactor cost?

    Science.gov (United States)

    Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

    2017-10-01

    The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

  11. Estimation of Zeff in Novillo Tokamak

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. 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

  13. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    Chen Xiangbo

    2001-01-01

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

  14. 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

  15. Power and particle exhaust in tokamaks

    International Nuclear Information System (INIS)

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER's nominal design positions; important directions for further research are identified

  16. Definition of total bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    Ross, D.W.

    1995-01-01

    Alternative definitions of the total bootstrap current are compared. An analogous comparison is given for the ohmic and auxiliary currents. It is argued that different definitions than those usually employed lead to simpler analyses of tokamak operating scenarios

  17. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

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

  18. Neutral beam in ALVAND IIC tokamak

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  19. 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

  20. Empirical scaling for present Ohmically heated tokamaks

    International Nuclear Information System (INIS)

    Daughney, C.

    1975-01-01

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

  1. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

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

  2. 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.

  3. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    Caldas, I.L.; Vannucci, A.

    1985-01-01

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt

  4. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  5. Theory of incremental turbulent transport in tokamaks

    International Nuclear Information System (INIS)

    Similon, P.L.

    1991-01-01

    The goal of this research is to understand how the various aspect of turbulent transport operate in tokamaks, in the presence of low frequency fluctuations such as drift waves or trapped electron modes

  6. Numerical simulation of edge plasma in tokamak

    International Nuclear Information System (INIS)

    Chen Yiping; Qiu Lijian

    1996-02-01

    The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)

  7. Plasma diagnostics using synchrotron radiation in tokamaks

    International Nuclear Information System (INIS)

    Fidone, I.; Giruzzi, G.; Granata, G.

    1995-09-01

    This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs

  8. 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)

  9. Electric conductivity and bootstrap current in tokamak

    International Nuclear Information System (INIS)

    Mao Jianshan; Wang Maoquan

    1996-12-01

    A modified Ohm's law for the electric conductivity calculation is presented, where the modified ohmic current can be compensated by the bootstrap current. A comparison of TEXT tokamak experiment with the theories shows that the modified Ohm's law is a more close approximation to the tokamak experiments than the classical and neoclassical theories and can not lead to the absurd result of Z eff <1, and the extended neoclassical theory would be not necessary. (3 figs.)

  10. Erosion of the first wall of Tokamaks

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  11. Advanced tokamak physics in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)

    2000-12-01

    Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)

  12. Optimization design for SST-1 Tokamak insulators

    International Nuclear Information System (INIS)

    Zhang Yuanbin; Pan Wanjiang

    2012-01-01

    With the help of ANSYS FEA technique, high voltage and cryogenic proper- ties of the SST-1 Tokamak insulators were obtained, and the structure of the insulators was designed and modified by taking into account the simulation results. The simulation results indicate that the optimization structure has better high voltage insulating property and cryogenic mechanics property, and also can fulfill the qualification criteria of the SST-1 Tokamak insulators. (authors)

  13. 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)

  14. Preliminary results of the TBR small tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Fagundes, A.N.; Da Silva, R.P.; Galvao, R.M.O.; Del Bosco, E.; Vuolo, J.H.; Sanada, E.K.; Dellaqua, R.

    1982-01-01

    The paper gives a short description of the TBR - small Brazilian tokamak and the first results obtained for plasma formation and equilibrium. Measured breakdown curves for hydrogen are shown to be confined within analytically calculated limits and to depend strongly on stray vertical magnetic fields. Time profiles of plasma current in equilibrium are shown and compared with the predictions of a simple analytical model for tokamak discharges. Reasonable agreement is obtained taking Zsub(eff) as a free parameter. (author)

  15. Full power in the European tokamak

    International Nuclear Information System (INIS)

    Lallia, P.P.; Hugon, M.

    1987-01-01

    A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years

  16. 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

  17. 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

  18. 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

  19. 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.

  20. Effects of q and high beta on tokamak stability

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  1. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    Pitcher, C.S.

    1987-11-01

    The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D 2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict

  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. D-D tokamak reactor assessment

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-03-01

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

  5. On the HL-1M tokamak plasma confinement time

    International Nuclear Information System (INIS)

    Qin Yunwen

    2001-01-01

    Emphasizing that the tokamak plasma confinement time is the plasma particle or thermal energy loss characteristic time, the relevant physical concept and HL-1M tokamak experimental data analyses are reviewed

  6. Formation and sustainment of a very low aspect ratio tokamak using coaxial helicity injection (the Helicity Injected [HIT] experiment). Annual progress report No. 5, December 1, 1993--December 31, 1994

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Nelson, B.A.

    1995-01-01

    This is the fifth Progress Report on the Helicity Injected Tokamak (HIT) at the University of Washington, Seattle, DOE Grant DE-FE06-90ER54095. This report covers the period of December 1, 1993 through December 31

  7. Analysis of tokamak plasma confinement modes using the fast

    Indian Academy of Sciences (India)

    The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...

  8. Comparative studies of stellarator and tokamak transport

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  9. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Furth, H.P.

    1985-10-01

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

  10. Edge plasma diagnostics on Tore Supra tokamak

    International Nuclear Information System (INIS)

    Fujita, Junji

    1991-01-01

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

  11. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J A

    1994-11-30

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

  12. Fast, multi-channel real-time processing of signals with microsecond latency using graphics processing units

    Energy Technology Data Exchange (ETDEWEB)

    Rath, N., E-mail: Nikolaus@rath.org; Levesque, J. P.; Mauel, M. E.; Navratil, G. A.; Peng, Q. [Department of Applied Physics and Applied Mathematics, Columbia University, 500 W 120th St, New York, New York 10027 (United States); Kato, S. [Department of Information Engineering, Nagoya University, Nagoya (Japan)

    2014-04-15

    Fast, digital signal processing (DSP) has many applications. Typical hardware options for performing DSP are field-programmable gate arrays (FPGAs), application-specific integrated DSP chips, or general purpose personal computer systems. This paper presents a novel DSP platform that has been developed for feedback control on the HBT-EP tokamak device. The system runs all signal processing exclusively on a Graphics Processing Unit (GPU) to achieve real-time performance with latencies below 8 μs. Signals are transferred into and out of the GPU using PCI Express peer-to-peer direct-memory-access transfers without involvement of the central processing unit or host memory. Tests were performed on the feedback control system of the HBT-EP tokamak using forty 16-bit floating point inputs and outputs each and a sampling rate of up to 250 kHz. Signals were digitized by a D-TACQ ACQ196 module, processing done on an NVIDIA GTX 580 GPU programmed in CUDA, and analog output was generated by D-TACQ AO32CPCI modules.

  13. Fast, multi-channel real-time processing of signals with microsecond latency using graphics processing units

    International Nuclear Information System (INIS)

    Rath, N.; Levesque, J. P.; Mauel, M. E.; Navratil, G. A.; Peng, Q.; Kato, S.

    2014-01-01

    Fast, digital signal processing (DSP) has many applications. Typical hardware options for performing DSP are field-programmable gate arrays (FPGAs), application-specific integrated DSP chips, or general purpose personal computer systems. This paper presents a novel DSP platform that has been developed for feedback control on the HBT-EP tokamak device. The system runs all signal processing exclusively on a Graphics Processing Unit (GPU) to achieve real-time performance with latencies below 8 μs. Signals are transferred into and out of the GPU using PCI Express peer-to-peer direct-memory-access transfers without involvement of the central processing unit or host memory. Tests were performed on the feedback control system of the HBT-EP tokamak using forty 16-bit floating point inputs and outputs each and a sampling rate of up to 250 kHz. Signals were digitized by a D-TACQ ACQ196 module, processing done on an NVIDIA GTX 580 GPU programmed in CUDA, and analog output was generated by D-TACQ AO32CPCI modules

  14. Neoclassical MHD equations for tokamaks

    International Nuclear Information System (INIS)

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

    1986-03-01

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

  15. 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)

  16. The collaborative tokamak control room

    International Nuclear Information System (INIS)

    Schissel, D.P.

    2006-01-01

    Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaborations between experimental sites and laboratories worldwide. In the US, the National Fusion Collaboratory Project is developing a persistent infrastructure to enable scientific collaboration for all aspects of magnetic fusion energy research by creating a robust, user-friendly collaborative environment and deploying this to the more than 1000 US fusion scientists in 40 institutions who perform magnetic fusion research. This paper reports on one aspect of the project which is the development of the collaborative tokamak control room to enhance both collocated and remote scientific participation in experimental operations. This work includes secured computational services that can be scheduled as required, the ability to rapidly compare experimental data with simulation results, a means to easily share individual results with the group by moving application windows to a shared display, and the ability for remote scientists to be fully engaged in experimental operations through shared audio, video, and applications. The project is funded by the USDOE Office of Science, Scientific Discovery through Advanced Computing (SciDAC) Program and unites fusion and computer science researchers to directly address these challenges

  17. Tokamak rotation and charge exchange

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  18. The Research Progress of the J-TEXT Tokamak

    Science.gov (United States)

    Zhuang, Ge; Wang, Zhijiang; Ding, Yonghua; Zhang, Ming; Yang, Zhoujun; Gao, Li; Zhang, Xiaoqing; Hu, Xiwei; Pan, Yuan

    2010-11-01

    In 2004, the TEXT-U tokamak was disassembled and shipped to China, and later on settle down in Huazhong University of Science and Technology. The machine was renamed as the Joint TEXT (J-TEXT) tokamak and obtained its first plasma in 2007. The typical J-TEXT Ohmic discharge was performed in the limiter configuration with the main parameters as follows: major radius R=1.05 m, minor radius a=0.27m, toroidal magnetic field BT=2.2T, plasma current Ip>200kA, line-averaged density ne˜ 2-3 . 1019/m^3, and electron temperature Te0˜ 700eV. Up till now, a few diagnostic systems used to facilitate routine operation and experimental studies were designed and developed. Benefiting from these diagnostic tools, the observation of MHD activities and the statistical analysis of disruption events were done. And measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the LCFS were also made as well. The preliminary results will be presented in detail in the meeting.

  19. Photometric Calibration of the SPRED at the FTU Tokamak

    International Nuclear Information System (INIS)

    May, M J

    1999-01-01

    The SPRED spectrometer was photometrically calibrated by using the FTU tokamak plasma and the Grazing Incidence Time Resolving Spectrometer (GRITS) from the Johns Hopkins University [Stratton, Nucl. Fusion, Vol. 24, No. 6, pp. 767-777, 1984]. The photometric calibration of the GRITS spectrometer was transferred to the SPRED [Fonck, R.J., Applied Optics, Vol. 21, No. 12, p. 2115 (1982)] by directly comparing the intensity of bright lines emitted from the FTU tokamak plasma that were simultaneously measured by both spectrometers. The GRITS spectrometer (λ = 10 - 360 (angstrom); Δλ ∼ 0.7 (angstrom)) was photometrically calibrated in the 50 - 360 (angstrom) spectral range at the SURF II synchrotron light source at NIST in Gaithersburg MD in August 1997. The calibration of each SPRED grating was performed separately. These gratings covered the short wavelengths: 100 - 300 (angstrom)(Δλ - 1.4 (angstrom)) and the long wavelengths: 200 - 1800 (angstrom) (Δλ ∼ 7 (angstrom)). This calibration should be accurate until the microchannel plate of the SPRED is exposed to atmospheric pressure. This calibration is similar to the one obtained by Stratton [Stratton, Rev. Sci. Instrum. 57 (8), pp. 204,3 August 1986

  20. 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.)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-09-01

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

  3. 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.)

  4. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

    Rebut, P.H.; Green, B.J.

    1977-01-01

    The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)

  5. 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

  6. Time - resolved thermography at Tokamak T-10

    International Nuclear Information System (INIS)

    Grunow, C.; Guenther, K.; Lingertat, J.; Chicherov, V.M.; Evstigneev, S.A.; Zvonkov, S.N.

    1987-01-01

    Thermographic experiments were performed at T-10 tokamak to investigate the thermic coupling of plasma and the limiter. The limiter is an internal equipment of the vacuum vessel of tokamak-type fusion devices and the interaction of plasma with limiter results a high thermal load of limiter for short time. In according to improve the limiter design the temperature distribution on the limiter surface was measured by a time-resolved thermographic method. Typical isotherms and temperature increment curves are presented. This measurement can be used as a systematic plasma diagnostic method because the limiter is installed in the tokamak whereas special additional probes often disturb the plasma discharge. (D.Gy.) 3 refs.; 7 figs

  7. 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.)

  8. 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

  9. Activation analysis of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

  10. Helicity content and tokamak applications of helicity

    International Nuclear Information System (INIS)

    Boozer, A.H.

    1986-05-01

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

  11. Atomic data for integrated tokamak modelling

    International Nuclear Information System (INIS)

    Toekesi, K.

    2013-01-01

    scattering. Moreover we present elastic cross sections of fusion related materials. We present total and angular differential elastic cross sections of hydrogen atoms for a wide range of incident electron energy. One of the convenient ways of representation of these data is analytical fit functions, which can be easily applied in various fields of sciences. The aim of work is to develop a universal functional formula of elastic cross sections for the case of hydrogen target. We consider the angular differential electron elastic cross sections for a wide range of incident electron energy and in the entire angular range. The differential cross-sections were calculated using the partial expansion method. The fitted curves are in excellent agreement with the calculated ones within less than 1% accuracy in the energy range between 1 eV and 100 keV. Our analytical formula may show the main virtues in various Monte Carlo simulations reducing drastically the computation time, when it requires to calculate the elastic cross sections many times. The applied fitting technique can be used for other data. Acknowledgement: This work, supported by the European Communities under the contract of Association between EURATOMHAS, was carried out within the framework of the Task Force on Integrated Tokamak Modeling of the European Fusion Development Agreement. The work was also supported by the Hungarian Scientific Research Fund OTKA No. NN103279. (author)

  12. 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.

  13. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

  14. Ballooning stable high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Tuda, Takashi; Azumi, Masafumi; Kurita, Gen-ichi; Takizuka, Tomonori; Takeda, Tatsuoki

    1981-04-01

    The second stable regime of ballooning modes is numerically studied by using the two-dimensional tokamak transport code with the ballooning stability code. Using the simple FCT heating scheme, we find that the plasma can locally enter this second stable regime. And we obtained equilibria with fairly high beta (β -- 23%) stable against ballooning modes in a whole plasma region, by taking into account of finite thermal diffusion due to unstable ballooning modes. These results show that a tokamak fusion reactor can operate in a high beta state, which is economically favourable. (author)

  15. Energy storage for tokamak reactor cycles

    International Nuclear Information System (INIS)

    Buchanan, C.H.

    1979-01-01

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

  16. Tokamak power systems studies, FY 1985

    International Nuclear Information System (INIS)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs

  17. 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

  18. Two-ion ICRF heating in Tokamaks

    International Nuclear Information System (INIS)

    Tennfors, E.

    1985-03-01

    The practical consequences for tokamak plasma heating in the ion cyclotron frequency regime of the two-dimensional treatment of the two-ion mode conversion layer are analyzed. The problem of evaluation of the condition for fast wave resonance is analyzed, as well as the limitations imposed by warm plasma effects. Simple ways to find the mode conversion surfaces when they exist are presented. Also for large tokamaks, it is possible to obtain mode conversion conditions for realistic antenna spectra provided species concentration and frequency are chosen such that the surface Epsilon = 0 intersects the plasma midplane just outside of the magnetic axis. (Author)

  19. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    Downum, W.B.

    1981-01-01

    The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

  20. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  1. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    Johner, J.

    1989-04-01

    The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aB t x of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

  2. 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

  3. Periodic disruptions in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Zoletnik, S.

    1988-11-01

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

  4. Gas blanket fueling of a tokamak reactor

    International Nuclear Information System (INIS)

    Gralnick, S.L.

    1978-01-01

    The purpose of this paper is a speculative investigation of the potential of fueling a Tokamak by introducing a sufficiently large quantity of gaseous deuterium and tritium at the vacuum wall boundary. It is motivated by two factors: current generation tokamaks are, in a manner of speaking, fueled from the edge quite successfully as is evidenced by pulse lengths that are long compared to particle recycling times, and by rapid plasma density increase produced by gas puffing, alternative, deep penetration fueling techniques that have been proposed possess severe technological problems and large costs

  5. Tokamak Engineering Technology Facility scoping study

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR

  6. Alfven wave heating in a tokamak reactor

    International Nuclear Information System (INIS)

    Borg, G.G.; Appert, K.; Knight, A.J.; Lister, J.B.; Vaclavik, J.

    1990-01-01

    A number of features of Alfven wave heating make it potentially attractive for use in large tokamak reactors. Among them are the availability and relativity low cost of the power supplies, the potential ability to act selectively on the current profile, and the probable absence of operational limits in size, fields or density. The physics of Alfven wave heating in a large tokamak is assessed. Present theoretical understanding of mode coupling and antenna loading is extrapolated to a large machine. The problem of a recessed antenna is analysed. Calculations of loading and discussion of various heating scenarios for the particular case of NET are also presented. (author). 23 refs, 18 figs, 4 tabs

  7. Can better modelling improve tokamak control?

    International Nuclear Information System (INIS)

    Lister, J.B.; Vyas, P.; Ward, D.J.; Albanese, R.; Ambrosino, G.; Ariola, M.; Villone, F.; Coutlis, A.; Limebeer, D.J.N.; Wainwright, J.P.

    1997-01-01

    The control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed. (author) 5 figs., 9 refs

  8. Increase in beta limit in tokamak plasmas

    International Nuclear Information System (INIS)

    Kamada, Yutaka

    2003-01-01

    This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)

  9. 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

  10. Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

  11. 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

  12. Tokamak startup with electron cyclotron heating

    International Nuclear Information System (INIS)

    Holly, D.J.; Prager, S.C.; Shepard, D.A.; Sprott, J.C.

    1980-04-01

    Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed

  13. Transformer Recharging with Alpha Channeling in Tokamaks

    International Nuclear Information System (INIS)

    Fisch, N.J.

    2009-01-01

    Transformer recharging with lower hybrid waves in tokamaks can give low average auxiliary power if the resistivity is kept high enough during the radio frequency (rf) recharging stage. At the same time, operation in the hot ion mode via alpha channeling increases the effective fusion reactivity. This paper will address the extent to which these two large cost saving steps are compatible.

  14. Criteria for initiation of tokamak disruptions

    International Nuclear Information System (INIS)

    Hopcraft, K.I.; Turner, M.F.

    1986-01-01

    The process by which a tokamak plasma evolves from an equilibrium state containing a saturated magnetic island to one which is disruptively unstable is discussed and illustrated by numerical simulation of a resistive magnetoplasma. Those elements which are required to initiate a disruption are delineated

  15. Feedback stabilization of axisymmetric modes in tokamaks

    International Nuclear Information System (INIS)

    Jardin, S.C.; Larrabee, D.A.

    1982-01-01

    Noncircular tokamak plasmas can be unstable to ideal MHD axisymmetric instabilities. Passive conductors with finite resistivity will at best slow down these instabilities to the resistive (L/R) time of the conductors. An active feedback system far from the plasma which responds on this resistive time can stabilize the system provided its mutual inductance with the passive coils is small enough

  16. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Scattering measurements in Tokamak type devices

    International Nuclear Information System (INIS)

    Matoba, Tohru

    1975-03-01

    Theories, experiments and proposals for light scattering in Tokamak type devices are reviewed. Thomson scattering, measuring method of the current density distribution by scattering and resonance fluorescence are summarily described. These methods may be useful for diagnosis of the fusion plasmas. The report may help planning of the measuring apparatus for the fusion plasmas in future. (auth.)

  18. Maintenance considerations of the STARFIRE commercial tokamak

    International Nuclear Information System (INIS)

    Trachsel, C.A.; Zahn, H.S.; Field, R.E.; Stevens, H.C.

    1979-01-01

    This paper presents the maintenance approach, the commercial tokamak design features that enhance maintenance and preliminary repair time and required mean-time-between-failures for major subsystems. Reactor hall building and maintenance equipment requirements including hot cells, coil rewinding, and cranes are discussed

  19. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    1978-08-01

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  20. Tokamak power plant burn cycle options

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1994-06-01

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

  1. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  2. Preliminary measurements on Tokamak KT-5

    International Nuclear Information System (INIS)

    Wen Yizhi; Wan Shude; Rong Furui; Haan Shengshen; Liu Wandong; Liu Lei

    1987-01-01

    A small tokamak, KT-5, has been put in to operation since 1984. The major and minor radius of the plasma are 30 and 4.5 cm, respectively. The parameters obtained in the first phase of KT-5 experiments are as follows B t = 0.45 T, I p ≥ 5 kA, q(α) σ = 50 eV

  3. Experimental methods to study tokamak plasma stability

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1978-01-01

    Experimental devices to measure external instability modes with small pick-up coils to detect poloidal magnetic field fluctuations, and internal modes with soft-X-ray detectors are discussed. The characteristics of these devices are calculated for a small tokamak (R 0 = 30 cm, a = 10 cm, I 0 50 KA). (author)

  4. MHD stability of vertically asymmetric tokamak equilibria

    International Nuclear Information System (INIS)

    Dalhed, H.E.; Grimm, R.C.; Johnson, J.L.

    1981-03-01

    The ideal MHD stability properties of a special class of vertically asymmetric tokamak equilibria are examined. The calculations confirm that no major new physical effects are introduced and the modifications can be understood by conventional arguments. The results indicate that significant departures from up-down symmetry can be tolerated before the reduction in β becomes important for reactor operation

  5. Computation of tokamak equilibria with steady flow

    International Nuclear Information System (INIS)

    Kerner, W.; Tokuda, Shinji

    1987-08-01

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

  6. 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)

  7. Anomalous periodic disruptions in tokamak plasma

    International Nuclear Information System (INIS)

    Montvai, A.; Tegze, M.; Valyi, I.

    1982-09-01

    Anomalously strong, periodic instabilities were observed in the MT-1 tokamak. Characteristics of these instabilities were partly similar to those of internal disruptions, but there were features making them different from the normal relaxational oscillations. Basic characteristics of the phenomenon were studied with the aid of generally used diagnostics. (author)

  8. Energy confinement of high-density tokamaks

    NARCIS (Netherlands)

    Schüller, F.C.; Schram, D.C.; Coppi, B.; Sadowski, W.

    1977-01-01

    Neoclassical ion heat conduction is the major energy loss mechanism in the center of an ohmically heated high-d. tokamak discharge (n>3 * 1020 m-3). This fixes the mutual dependence of plasma quantities on the axis and leads to scaling laws for the poloidal b and energy confinement time, given the

  9. Supravodivý tokamak dobyl Asii

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2006-01-01

    Roč. 54, č. 18 (2006), s. 58 ISSN 0040-1064 Institutional research plan: CEZ:AV0Z20430508 Keywords : superconducting tokamak * ITER * Tore Supra * Institute of Plasma Physics AV CR Subject RIV: BL - Plasma and Gas Discharge Physics

  10. Runaway electrons in the TRIAM-1 tokamak

    International Nuclear Information System (INIS)

    Satoh, Takemichi; Nakamura, Kazuo; Toi, Kazuo; Nakamura, Yukio; Hiraki, Naoji

    1981-01-01

    Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges. (author)

  11. Runaway electrons in the TRIAM-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-01

    Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges.

  12. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

    also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...

  13. Central control system for the EAST tokamak

    International Nuclear Information System (INIS)

    Sun Xiaoyang; Ji Zhenshan; Wu Yicun; Luo Jiarong

    2008-01-01

    The architecture, the main function and the design scheme of the central control system and the collaboration system of EAST tokamak are described. The main functions of the central control system are to supply a union control interface for all the control, diagnoses, and data acquisition (DAQ) subsystem and it is also designed to synchronize all those subsystem. (authors)

  14. Microinstabilities in weak density gradient tokamak systems

    International Nuclear Information System (INIS)

    Tang, W.M.; Rewoldt, G.; Chen, L.

    1986-04-01

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient

  15. Design parameters of Tokamak-7 system

    International Nuclear Information System (INIS)

    Ivanov, D.P.; Keilin, V.E.; Klimenko, E.Yu.; Strelkov, V.S.

    Superconducting windings for the main magnetic field of Tokamak-7 are discussed. The parameters of this facility are based on the use of commercially available superconducting materials for fields up to 80 kOe. Experimental parameters are described. (U.S.)

  16. Modular pulse sequencing in a tokamak system

    International Nuclear Information System (INIS)

    Chew, A.C.; Lee, S.; Saw, S.H.

    1992-01-01

    Pulse technique applied in the timing and sequencing of the various part of the MUT tokamak system are discussed. The modular architecture of the pulse generating device highlights the versatile application of the simple physical concepts in precise and complicated research experiment. (author)

  17. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    HUMPHREYS, D.A.; FERRON, J.R.; JOHNSON, R.D; LEUER, J.A.; PENAFLOR, B.G.; WALKER, M.L.; WELANDER, A.S.; KHAYRUTDINOV, R.R; DOKOUKA, V.; EDGELL, D.H.; FRANSSON, C.M.

    2004-03-01

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance

  18. Tokamak startup with electron cyclotron heating

    Energy Technology Data Exchange (ETDEWEB)

    Holly, D J; Prager, S C; Shepard, D A; Sprott, J C

    1980-04-01

    Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed.

  19. 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

  20. 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

  1. 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)

  2. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  3. Parallel gradient effects on ICRH (Ion Cyclotron Resonance Heating) in Tokamaks

    International Nuclear Information System (INIS)

    Smithe, D.N.

    1987-01-01

    This dissertation examines the effects on Ion Cyclotron Resonance Heating of parallel nonuniformity in the magnetic field which arises from the poloidal field in a tokamak and the universal (major radius)/sup /minus/1/ scaling of the cyclotron frequency. The goal of the analysis is the macroscopic warm plasma current including temperature in the sense of the finite Larmor radius expansion and the quasilocal approximation of the parallel guiding center motion. A 1-D numerical application of the fully nonlocal integral dielectric is performed. Parallel gradient effects are studied for He-3 minority, 2nd harmonic deuterium, and hydrogen minority heating in tokamaks. The results show quite significant alteration of the toroidal wavenumber absorption spectrum, and a wealth of new behavior on the local propagation scale. 95 refs., 37 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. Influence of the helical resonant fields on the plasma potential in the TBR-1 Tokamak

    International Nuclear Information System (INIS)

    Ribeiro, C.; Silva, R.P. da; Caldas, I.L.; Fagundes, A.N.; Sanada, E.K.

    1990-01-01

    This work describes an experimental work that are in progress in TBR-1 tokamak about the influence of resonant helical fields on the plasma potential. TBR-1 is a small tokamak in operation in the Physics Institute of University of Sao Paulo and used for basic research, diagnostic development and personal formation. Its main parameters are: R(Major Radius) = 0.30 m; a v (Vessel Radius) = 0.11 m; a(Plasma Radius) = 0.08 m; R/a(Aspect Ratio) = 3.75; B φ (Toroidal Field) = 5 kG; n e0 (Central Electron Density) ≅ 7 x 10 18 m -3 ; T e0 (central electron temperature) ≅ 200 eV. (Author)

  7. An overview on plasma disruption mitigation and avoidance in tokamak

    International Nuclear Information System (INIS)

    He Kaihui; Pan Chuanhong; Feng Kaiming

    2002-01-01

    Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

  8. 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.

  9. Disruption generated secondary runaway electrons in present day tokamaks

    International Nuclear Information System (INIS)

    Pankratov, I.M.; Jaspers, R.

    2000-01-01

    An analysis of the runaway electron secondary generation during disruptions in present day tokamaks (JET, JT-60U, TEXTOR) was made. It was shown that even for tokamaks with the plasma current I approx 100 kA the secondary generation may dominate the runaway production during disruptions. In the same time in tokamaks with I approx 1 MA the runaway electron secondary generation during disruptions may be suppressed

  10. 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)

  11. 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

  12. Characterizing electrostatic turbulence in tokamak plasmas with high MHD activity

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes-Filho, Z O; Santos Lima, G Z dos; Caldas, I L; Nascimento, I C; Kuznetsov, Yu K [Instituto de Fisica, Universidade de Sao Paulo, Caixa Postal 66316, 05315-970, Sao Paulo, SP (Brazil); Viana, R L, E-mail: viana@fisica.ufpr.b [Departamento de Fisica, Universidade Federal do Parana, Caixa Postal 19044, 81531-990, Curitiba, PR (Brazil)

    2010-09-01

    One of the challenges in obtaining long lasting magnetic confinement of fusion plasmas in tokamaks is to control electrostatic turbulence near the vessel wall. A necessary step towards achieving this goal is to characterize the turbulence level and so as to quantify its effect on the transport of energy and particles of the plasma. In this paper we present experimental results on the characterization of electrostatic turbulence in Tokamak Chauffage Alfven Bresilien (TCABR), operating in the Institute of Physics of University of Sao Paulo, Brazil. In particular, we investigate the effect of certain magnetic field fluctuations, due to magnetohydrodynamical (MHD) instabilities activity, on the spectral properties of electrostatic turbulence at plasma edge. In some TCABR discharges we observe that this MHD activity may increase spontaneously, following changes in the edge safety factor, or after changes in the radial electric field achieved by electrode biasing. During the high MHD activity, the magnetic oscillations and the plasma edge electrostatic turbulence present several common linear spectral features with a noticeable dominant peak in the same frequency. In this article, dynamical analyses were applied to find other alterations on turbulence characteristics due to the MHD activity and turbulence enhancement. A recurrence quantification analysis shows that the turbulence determinism radial profile is substantially changed, becoming more radially uniform, during the high MHD activity. Moreover, the bicoherence spectra of these two kinds of fluctuations are similar and present high bicoherence levels associated with the MHD frequency. In contrast with the bicoherence spectral changes, that are radially localized at the plasma edge, the turbulence recurrence is broadly altered at the plasma edge and the scrape-off layer.

  13. Summary of the 1982 small tokamak users meeting

    International Nuclear Information System (INIS)

    Sprott, J.C.

    1982-11-01

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

  14. 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

  15. 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

  16. Assembly study for JT-60SA tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  17. 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)

  18. Neural net prediction of tokamak plasma disruptions

    International Nuclear Information System (INIS)

    Hernandez, J.V.; Lin, Z.; Horton, W.; McCool, S.C.

    1994-10-01

    The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system

  19. 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

  20. 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

  1. 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

  2. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.

    2011-01-01

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

  3. The tokamak - an imperfect frame of refernce

    International Nuclear Information System (INIS)

    Schmitter, K.U.

    1981-03-01

    It is attempted to assess the suitability of tokamaks for fusion power plants on the basis of existing design studies by reference to the reality of energy production in fission power plants. A definition of suitability criteria and a discussion of their relation to the most important features of power plants are followed by a comparative treatment. For example, the mean volumetric net electric power density in the nuclear islands of tokamak power plant designs is only 2,5 to 4 E of the value common today in light water reactor nuclear islands. In addition, configuration problems, auxiliary power requirements and energy payback time are discussed and taken into account in the assessment. (orig.)

  4. 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

  5. Microinstability theory in tokamaks: a review

    International Nuclear Information System (INIS)

    Tang, W.M.

    1977-06-01

    Significant investigations in the area of tokamak microinstability theory are reviewed. Emphasis is given to the work covering the period from 1970 through 1976. Special attention is focused on low-frequency electrostatic drift-type modes, which are generally believed to be the dominant tokamak microinstabilities under normal operating conditions. The basic linear formalism including electromagnetic (finite beta) modifications is presented along with a general survey of the numerous papers investigating specific linear and nonlinear effects on these modes. Estimates of the associated anomalous transport and confinement times are discussed, and a summary of relevant experimental results is given. Studies of the nonelectrostatic and high-frequency instabilities associated with the presence of high energy ions from neutral beam injection (or with the presence of alpha particles from fusion reactions) are also surveyed

  6. 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)

  7. Plasma internal inductance dynamics in a tokamak

    International Nuclear Information System (INIS)

    Romero, J.A.

    2010-01-01

    A lumped parameter model for tokamak plasma current and inductance time evolution as a function of plasma resistance, non-inductive current drive sources and boundary voltage or poloidal field coil current drive is presented. The model includes a novel formulation leading to exact equations for internal inductance and plasma current dynamics. Having in mind its application in a tokamak inductive control system, the model is expressed in state space form, the preferred choice for the design of control systems using modern control systems theory. The choice of system states allows many interesting physical quantities such as plasma current, inductance, magnetic energy, and resistive and inductive fluxes be made available as output equations. The model is derived from energy conservation theorem, and flux balance theorems, together with a first order approximation for flux diffusion dynamics. The validity of this approximation has been checked using experimental data from JET showing an excellent agreement.

  8. Runaway acceleration during magnetic reconnection in tokamaks

    International Nuclear Information System (INIS)

    Helander, P; Eriksson, L-G; Andersson, F

    2002-01-01

    In this paper, the basic theory of runaway electron production is reviewed and recent progress is discussed. The mechanisms of primary and secondary generation of runaway electrons are described and their dynamics during a tokamak disruption is analysed, both in a simple analytical model and through numerical Monte Carlo simulation. A simple criterion for when these mechanisms generate a significant runaway current is derived, and the first self-consistent simulations of the electron kinetics in a tokamak disruption are presented. Radial cross-field diffusion is shown to inhibit runaway avalanches, as indicated in recent experiments on JET and JT-60U. Finally, the physics of relativistic post-disruption runaway electrons is discussed, in particular their slowing down due to emission of synchrotron radiation, and their ability to produce electron-positron pairs in collisions with bulk plasma ions and electrons

  9. Design and construction of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.

    2001-01-01

    The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and 'advanced tokamak' physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long-pulse current-drive and heating systems. The 'advanced tokamak' aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. (author)

  10. Currents in the DIII-D Tokamak

    Science.gov (United States)

    Azari, A.; Eidietis, N. W.

    2012-10-01

    Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

  11. 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

  12. Microinstability theory in tokamaks: a review

    Energy Technology Data Exchange (ETDEWEB)

    Tang, W.M.

    1977-06-01

    Significant investigations in the area of tokamak microinstability theory are reviewed. Emphasis is given to the work covering the period from 1970 through 1976. Special attention is focused on low-frequency electrostatic drift-type modes, which are generally believed to be the dominant tokamak microinstabilities under normal operating conditions. The basic linear formalism including electromagnetic (finite beta) modifications is presented along with a general survey of the numerous papers investigating specific linear and nonlinear effects on these modes. Estimates of the associated anomalous transport and confinement times are discussed, and a summary of relevant experimental results is given. Studies of the nonelectrostatic and high-frequency instabilities associated with the presence of high energy ions from neutral beam injection (or with the presence of alpha particles from fusion reactions) are also surveyed.

  13. Boundary Plasma Turbulence Simulations for Tokamaks

    International Nuclear Information System (INIS)

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

    2008-05-01

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

  14. Models for impurity effects in tokamaks

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1980-03-01

    Models for impurity effects in tokamaks are described with an emphasis on the relationship between attainment of high β and impurity problems. We briefly describe the status of attempts to employ neutral beam heating to achieve high β in tokamaks and propose a qualitative model for the mechanism by which heavy metal impurities may be produced in the startup phase of the discharge. We then describe paradoxes in impurity diffusion theory and discuss possible resolutions in terms of the effects of large-scale islands and sawtooth oscillations. Finally, we examine the prospects for the Zakharov-Shafranov catastrophe (long time scale disintegration of FCT equilibria) in the context of present and near-term experimental capability

  15. 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)

  16. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-09-01

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

  17. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-06-01

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

  18. Constrained ripple optimization of Tokamak bundle divertors

    International Nuclear Information System (INIS)

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

    1983-02-01

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

  19. A review of lifetime analyses for tokamaks

    International Nuclear Information System (INIS)

    Harkness, S.D.; Cramer, B.

    1979-01-01

    System studies have shown that economic fusion power can best be achieved from the use of long lived components. The stresses generated in a first wall module are a complex function of its geometry, the chosen structural material and the tokamak burn cycle characteristics. A means of applying ASME Code Case 1592 to preliminary design has been established. Methods of incorporating some of the material property changes expected from irradiation are discussed. Cyclic stresses imposed by tokamak operation are expected to cause fatigue related properties to govern the life of the structure. Stress assisted bubble growth is also discussed. This may be the critical mechanism in establishing the creep rupture life of a fusion first wall component. (orig.)

  20. Decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives

  1. 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).

  2. Resonant helical fields in the TBR tokamak

    International Nuclear Information System (INIS)

    Bender, O.W.

    1986-01-01

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

  3. User's manual of Tokamak Simulation Code

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  4. 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

  5. Electron cyclotron emission from the PLT tokamak

    International Nuclear Information System (INIS)

    Hosea, J.; Arunasalam, V.; Cano, R.

    1977-07-01

    Experimental measurements of electron cyclotron emission from the PLT tokamak plasma reveal that black-body emission occurs at the fundamental frequency. Such emission, not possible by direct thermal excitation of electromagnetic waves, is herein attributed to thermal excitation of electrostatic (Bernstein) waves which then mode convert into electromagnetic waves. The local feature of the electrostatic wave generation permits spatially and time resolved measurements of electron temperature as for the second harmonic emission

  6. Discrete Alfven waves in the TORTUS tokamak

    International Nuclear Information System (INIS)

    Amagishi, Y.; Ballico, M.J.; Cross, R.C.; Donnely, I.J.

    1989-01-01

    Discrete Alfven Waves (DAWs) have been observed as antenna resistance peaks and as enhanced edge fields in the TORTUS tokamak during Alfven wave heating experiments. A kinetic theory code has been used to calculate the antenna loading and the structure of the DAW fields for a range of plasma current and density profiles. There is fair agreement between the measured and predicted amplitude of the DAW fields in the plasma edge when both are normalized to the same antenna power

  7. 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)

  8. 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)

  9. Axisymmetric instability in a noncircular tokamak

    International Nuclear Information System (INIS)

    Lipschultz, B.

    1979-10-01

    The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes - the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria

  10. Neoclassical tearing modes in a tokamak

    International Nuclear Information System (INIS)

    Hahm, T.S.

    1988-08-01

    Linear tearing instability is studied in the banana collisionality regime in tokamak geometry. Neoclassical effects produce significant modifications of Ohm's law and the vorticity equation so that the growth rate of tearing modes driven by Δ' is dramatically reduced compared to the usual resistive MHD value. Consequences of this result, regarding the presence of pressure-gradient-driven neoclassical resistive interchange instabilities and the evolution of magnetic islands in the Rutherford regime, are discussed. 10 refs

  11. 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#

  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. Neutral-beam current drive in tokamaks

    International Nuclear Information System (INIS)

    Devoto, R.S.

    1986-01-01

    The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500 to 700 keV are needed for this device

  14. Neutral-beam current drive in tokamaks

    International Nuclear Information System (INIS)

    Devoto, R.S.

    1987-01-01

    The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500-700 keV are needed for this device

  15. Confinement scaling and ignition in tokamaks

    International Nuclear Information System (INIS)

    Perkins, F.W.; Sun, Y.C.

    1985-10-01

    A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 10 15 cm -3 , high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition

  16. Comparison of tokamak burn cycle options

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commerical reactor as well as an INTOR-class device. We begin with a review of the burn cycle options

  17. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  18. 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

  19. High beta plasmas in the PBX tokamak

    International Nuclear Information System (INIS)

    Bol, K.; Buchenauer, D.; Chance, M.

    1986-04-01

    Bean-shaped configurations favorable for high β discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present β limit

  20. 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

  1. Small-scale tearing mode in tokamaks

    International Nuclear Information System (INIS)

    Ivanov, N.V.

    1983-01-01

    Considerations are given on the possible effect of small-scale tearing mode with m >> 1 on the plasma electron thermal conductivity in a tokamak. The estimate of the electron thermal conductivity coefficient is obtained. Calculation results are compared with experimental data. The calculated dependence of radial distribution of electron temperature is shown to vary weakly with the tn(m 2 /m 1 ) alteration everywhere, except for the vicinity of point r approximately 0

  2. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2005-01-01

    SST-1, a steady state superconducting tokamak, is undergoing commissioning tests at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in a tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel. 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. The auxiliary current drive is based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. Detailed commissioning tests on the cryogenic system and experiments on the hydraulic characters and cool down features of single TF coils have been completed prior to the cool down of the entire superconducting system. Results of the single TF magnet cool down, and testing of the magnet system are presented. First experiments related to the breakdown and the current ramp up will subsequently be carried out. (author)

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

    International Nuclear Information System (INIS)

    Azizov, E A

    2012-01-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

  4. 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.

  5. 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)

  6. 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.

  7. Accessibility of high β tokamak states

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1978-05-01

    Encouraging results with neutral beam heating and adiabatic compression of tokamak plasmas have prompted new experiments which will study the approach to high β states. As projected tokamak β values become nonnegligible (average β of 4% is the goal), the models previously used for transport calculations will become inadequate. These models will be required to account for the evolution of the magnetic geometry, along with the change in plasma parameters. We present an axisymmetric transport model which should be useful for studying the approach to higher β values in tokamak experiments. Results from transport calculations with this model allow us to draw a parallel between observed behavior in seemingly unrelated experiments: electron heating by neutral injection in the ORMAK device and adiabatic compression in the ATC experiment. Finally, we find that the nature of cross-field transport may be expected to change as significant β values are reached. Enhanced transport from ballooning instabilities is likely to play a role as important as that now played by sawtooth (m = 1) and saturated (m = 2) instabilities. New techniques for describing this transport are required

  8. 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.

  9. Relativistic runaway electrons in tokamak plasmas

    International Nuclear Information System (INIS)

    Jaspers, R.E.

    1995-01-01

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

  10. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

    2015-10-15

    Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

  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. Super high field ohmically heated tokamak operation

    International Nuclear Information System (INIS)

    Cohn, D.R.; Bromberg, L.; Leclaire, R.J.; Potok, R.E.; Jassby, D.L.

    1986-01-01

    The authors discuss a super high field mode of tokamak operation that uses ohmic heating or near ohmic heating to ignition. The super high field mode of operation uses very high values of Β/sup 2/α, where Β is the magnetic field and a is the minor radius (Β/sup 2/α > 100 T/sup 2/m). We analyze copper magnet devices with major radii from 1.7 to 3.0 meters. Minimizing or eliminating the need for auxiliary heating has the potential advantages of reducing uncertainty in extrapolating the energy confinement time of current tokamak devices, and reducing engineering problems associated with large auxiliary heating requirements. It may be possible to heat relatively short pulse, inertially cooled tokamaks to ignition with ohmic power alone. However, there may be advantages in using a very small amount of auxiliary power (less than the ohmic heating power) to boost the ohmic heating and provide a faster start-up, expecially in relatively compact devices

  13. The ICRH tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1976-01-01

    A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

  14. On the density limit of Tokamaks

    International Nuclear Information System (INIS)

    Lehnert, B.

    1982-12-01

    Under the conditions of so far performed quasi-steady tokamak experiments near the density limit, the plasma pressure gradient in the outer layers of the plasma body becomes mainly determined by the plasma-neutral gas balance. An earlier analysis of ballooning instabilities driven by this gradient in regions of bad curvature has been extended to deduce an explicit stability criterion which determines the density limit. This criterion is closely related to the empirical Murakami limit. At relevant tokamak data, the deduced limit becomes proportional to J(sub)zR(sup)1/2 where J(sub)z is the average current density and R the major plasma radius. It is further found to be independent of the toroidal magnetic field strength and anomalous transport, as well as to be a slow function of the outer layer temperature and the mass number. The deduced stability criterion is consistent with so far performed experiments. Provided that the present analysis can be extrapolated to a wider range of parameter data and be combined with Alcator scaling, conditions near ignition appear to become realizable in small tokamaks by ohmic heating alone. These conditions can be satisfied at relevant magnetic field strengths and plasma currents, by imposing a high plasma current density. (author)

  15. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    Chan, V.S.

    1994-09-01

    In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

  16. The design of the KSTAR tokamak

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. 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

  18. Deposit of thin films for Tokamaks conditioning

    International Nuclear Information System (INIS)

    Valencia A, R.

    2006-01-01

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature ( -6 to 4.5 x 10 -6 Ω-m, thus taking the Z ef value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow

  19. A multimode analytic cylindrical model for the stabilization of the resistive wall modes

    International Nuclear Information System (INIS)

    Miron, I G

    2008-01-01

    A dispersion relation concerning the stability of the resistive wall modes within a multimode cylindrical analytical model is presented. This paper generalizes the Fitzpatrick-Aydemir model (Fitzpatrick R and Aydemir A Y 1996 Nucl. Fusion 1 11) in the presence of an unlimited number of neighboring modes for a tokamak plasma column surrounded by a resistive shell and a feedback system consisting of a number of detector and active feedback coils. The model is applied to the HBT-EP tokamak (Cates C et al 2000 Phys. Plasmas 7 3133) with its peculiar feedback system disposal. Finally, an analytical dispersion relation is obtained that can be solved by using a simple MATLAB code

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

    Science.gov (United States)

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

    2011-09-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  2. Research tokamak system with multi-mode discharges using inverter power supply

    International Nuclear Information System (INIS)

    Kojima, Hiroki; Kobayashi, Masahiro; Takagi, Makoto; Takamura, Shuichi; Tashiro, Kenji

    1999-01-01

    In Current Sustaining Tokamak in Nagoya university (CSTN)-IV research tokamak system using a compact 40kHz pulse width modulation (PWM) inverter power supply, which is controlled through LabVIEW program, we construct a new tokamak discharge system with multi-mode including a stable alternating current discharge and a high-repetition high-duty one. These discharge modes can be operated continuously for as long as 60sec. The continuous discharge with long duration is able to simulate the important physical and chemical processes of long time discharges in fusion devices, in which the heat load to the wall and the particle balance in the plasma-wall system are crucial topics in order to realize a long pulse fusion reactor, like ITER. Employing ergodic divertor (ED) is one of tools to control the particle balance and the heat load to the wall. In addition, we installed another inverter power supply to generate a rotating magnetic perturbation for dynamic ergodic divertor (DED) with the appropriate measurement system so that we may carry out experiments on heat and particle control with DED at long time operation. (author)

  3. Recent progress on the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    Ignat, D.W.

    1987-01-01

    This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule

  4. Tokamak WEST připraven ke startu!

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Květen (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * WEST * Tora Supra * divertor Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/2014-tokamak-west-pripraven-ke- start u

  5. 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.

  6. Commercial feasibility of fusion power based on the tokamak concept

    International Nuclear Information System (INIS)

    Reid, R.L.; Steiner, D.

    1977-01-01

    The impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants is determined. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  7. Lower hybrid heating experiments in tokamaks: an overview

    International Nuclear Information System (INIS)

    Porkolab, M.

    1985-10-01

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

  8. 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)

  9. 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

  10. Physics design requirements for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

    Neilson, G.H.; Goldston, R.J.; Jardin, S.C.; Reiersen, W.T.; Porkolab, M.; Ulrickson, M.

    1993-01-01

    The design of TPX is driven by physics requirements that follow from its mission. The tokamak and heating systems provide the performance and profile controls needed to study advanced steady state tokamak operating modes. The magnetic control systems provide substantial flexibility for the study of regimes with high beta and bootstrap current. The divertor is designed for high steady state power and particle exhaust

  11. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

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

  12. Design of Tokamak plasma with high Tc superconducting coils

    International Nuclear Information System (INIS)

    Uchimoto, T.; Miya, K.; Yoshida, Y.; Yamada, T.

    1999-01-01

    This paper presents a design of tokamak plasma in light of how the small ignited tokamak is possible with use of the HTSC coils as plasma stabilizer. The same data base and formulas as ITER are here used and any innovative technology other than the HTSC stabilizing coils is not assumed. (author)

  13. 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)

  14. 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

  15. High performance operational limits of tokamak and helical systems

    International Nuclear Information System (INIS)

    Yamazaki, Kozo; Kikuchi, Mitsuru

    2003-01-01

    The plasma operational boundaries of tokamak and helical systems are surveyed and compared with each other. Global confinement scaling laws are similar and gyro-Bohm like, however, local transport process is different due to sawtooth oscillations in tokamaks and ripple transport loss in helical systems. As for stability limits, achievable tokamak beta is explained by ideal or resistive MHD theories. On the other hand, beta values obtained so far in helical system are beyond ideal Mercier mode limits. Density limits in tokamak are often related to the coupling between radiation collapse and disruptive MHD instabilities, but the slow radiation collapse is dominant in the helical system. The pulse length of both tokamak and helical systems is on the order of hours in small machines, and the longer-pulsed good-confinement plasma operations compatible with radiative divertors are anticipated in both systems in the future. (author)

  16. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.

    1993-01-01

    A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

  17. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. The physics of an ignited tokamak

    International Nuclear Information System (INIS)

    Troyon, F.

    1990-10-01

    There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab

  20. Design of the ITER Tokamak Assembly Tools

    International Nuclear Information System (INIS)

    Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

    2006-01-01

    ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

  1. Dynamics of fast ions in Tokamaks

    International Nuclear Information System (INIS)

    Helander, P.

    1994-01-01

    Fast ions play a prominent role in the heating of tokamak plasmas by, e.g. neutral-beam injection, ion-cyclotron-resonance heating, and alpha-particle heating. In this thesis, a number of physical and mathematical problems concerning the dynamics of fast ions in tokamaks are addressed. First, the motion under adiabatic perturbations is studied. The frequencies of instabilities excited in tokamaks sometimes vary slowly with time. The existence of an adiabatic invariant of particle motion in such circumstances is shown to lead to a rapid convection of particles in the radial direction. Generalized adiabatic invariants are constructed for systems where the slowly varying parameter is subjected to small, but rapidly varying, fluctuations. Second, the onset of stochastic motion under resonant perturbations is considered. It is shown that the finite width of fast-ion drift orbits significantly affects the threshold for stochastic motion caused by magnetic field ripple or ion-cyclotron-resonance heating. Finite-orbit-width effects are also shown to reduce the strength of resonant interaction between alpha particles and internal kink modes. Third, the diffusive motion in the stochastic regime is analysed mathematically. Monte Carlo operators for the motion on long time-scales are constructed, and the validity of the quasilinear diffusion coefficient is examined. Finally, the effects of close ion collisions are investigated. It is demonstrated that close encounters with fast ions produce a high-energy tail in the distribution functions of impurity ions, and that close collisions between fusion-generated alpha particles give rise to a population of such particles with energies extending up to twice the birth energy. 44 refs

  2. Mathematical modeling plasma transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)

    1997-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 1020/m3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

  3. Mathematical modeling plasma transport in tokamaks

    International Nuclear Information System (INIS)

    Quiang, Ji

    1995-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

  4. Neutral beam injection on the PLT tokamak

    International Nuclear Information System (INIS)

    Schilling, G.; Ashcroft, D.L.; Eubank, H.P.; Grisham, L.R.; Knauer, R.C.; Stewart, L.D.; Stooksberry, R.W.; Ulrickson, M.; Williams, M.D.

    1981-01-01

    We describe the operation of the neutral beam injection system on the PLT tokamak. Improvements, retrofits, and conditioning have changed the injection system from an experiment in itself to a fairly reliable and useful plasma heating tool. We will present a brief overview of our physics achievements and then describe the system as it exists now. This will include injector performance, conditioning needs, maintenance needs, reliability, and daily operating sequences. We will also include hardware modifications and additions, electrical and mechanical, and point out remaining problem areas

  5. The calculation of Tritium burnup in Tokamaks

    International Nuclear Information System (INIS)

    Bittoni, E.; Haegi, M.

    1987-01-01

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

  6. The steady-state tokamak program

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  7. 'Crescent'-shaped tokamak for compact ignition

    International Nuclear Information System (INIS)

    Yamazaki, K.; Reiersen, W.T.

    1985-12-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross-section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design dose not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  8. Modification of tokamak edge turbulence using feedback

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  9. Particle and heat transport in Tokamaks

    International Nuclear Information System (INIS)

    Chatelier, M.

    1984-01-01

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

  10. 'Crescent'-shaped tokamak for compact ignition

    International Nuclear Information System (INIS)

    Yamazaki, K.; Reiersen, W.T.

    1986-01-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design does not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  11. Filamentary probe on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Kovařík, Karel; Ďuran, Ivan; Stöckel, Jan; Seidl, Jakub; Adámek, Jiří; Spolaore, M.; Vianello, N.; Háček, Pavel; Hron, Martin; Pánek, Radomír

    2017-01-01

    Roč. 88, č. 3 (2017), č. článku 035106. ISSN 0034-6748 R&D Projects: GA MŠk(CZ) 8D15001; GA ČR(CZ) GA15-10723S; GA ČR(CZ) GA16-25074S Institutional support: RVO:61389021 Keywords : tokamak * filaments * scrape-off layer Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.515, year: 2016 http://aip.scitation.org/doi/10.1063/1.4977591

  12. 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)

  13. Dust remobilization experiments on the COMPASS tokamak.

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Matějíček, Jiří; Ratynskaia, S.; Tolias, P.; De Angeli, M.; Riva, G.; Dimitrova, Miglena; Havlíček, Josef; Adámek, Jiří; Seidl, Jakub; Tomeš, Matěj; Cavalier, Jordan; Imríšek, Martin; Havránek, Aleš; Pánek, Radomír; Peterka, Matěj

    2017-01-01

    Roč. 124, November (2017), s. 446-449 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : Dust remobilization * Tungsten * Disruption * ELM * Plasma * Tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617300650

  14. Tokamak physics experiment: Diagnostic windows study

    International Nuclear Information System (INIS)

    Merrigan, M.; Wurden, G.A.

    1995-11-01

    We detail the study of diagnostic windows and window thermal stress remediation in the long-pulse, high-power Tokamak Physics Experiment (TPX) operation. The operating environment of the TPX diagnostic windows is reviewed, thermal loads on the windows estimated, and cooling requirements for the windows considered. Applicable window-cooling technology from other fields is reviewed and its application to the TPX windows considered. Methods for TPX window thermal conditioning are recommended, with some discussion of potential implementation problems provided. Recommendations for further research and development work to ensure performance of windows in the TPX system are presented

  15. Viscosity in the edge of tokamak plasmas

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1993-05-01

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

  16. Iron forbidden lines in tokamak discharges

    International Nuclear Information System (INIS)

    Suckewer, S.; Hinnov, E.

    1979-03-01

    Several spectrum lines from forbidden transitions in the ground configurations of highly ionized atoms have been observed in the PLT tokamak discharges. Such lines allow localized observations, in the high-temperature regions of the plasma, of ion-temperatures, plasma motions, and spatial distributions of ions. Measured absolute intensities of the forbidden lines have been compared with simultaneous observations of the ion resonance lines and with model calculations in order to deduce the mechanism of level populaions by means of electron collisions and radiative transitions

  17. Iron forbidden lines in tokamak discharges

    Energy Technology Data Exchange (ETDEWEB)

    Suckewer, S.; Hinnov, E.

    1979-03-01

    Several spectrum lines from forbidden transitions in the ground configurations of highly ionized atoms have been observed in the PLT tokamak discharges. Such lines allow localized observations, in the high-temperature regions of the plasma, of ion-temperatures, plasma motions, and spatial distributions of ions. Measured absolute intensities of the forbidden lines have been compared with simultaneous observations of the ion resonance lines and with model calculations in order to deduce the mechanism of level populaions by means of electron collisions and radiative transitions.

  18. DIII-D Advanced Tokamak Research Overview

    International Nuclear Information System (INIS)

    V.S. Chan; C.M. Greenfield; L.L. Lao; T.C. Luce; C.C. Petty; G.M. Staebler

    1999-01-01

    This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously β N H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues

  19. 'Snowflake' H Mode in a Tokamak Plasma

    International Nuclear Information System (INIS)

    Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.

    2010-01-01

    An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a 'snowflake' (second-order null) divertor configuration for the first time in the TCV tokamak. This regime exhibits 2 to 3 times lower ELM frequency and 20%-30% increased normalized ELM energy (ΔW ELM /W p ) compared to an identically shaped, conventional single-null diverted H mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.

  20. Structural features of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Citrolo, J.; Brown, G.; Rogoff, P.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is undergoing preliminary structural design and definitions. It will be relatively inexpensive with ignition capabilities. During the definition phase it was concluded that the TF coil should be assembled from the laminate copper-Inconel plates since copper alone cannot sustain the expected magnetic and thermal loads. An extensive test program is being initiated to investigate the various materials, and their elastic and inelastic response and to develop the constitutive equations required for the selection of design criteria and for the stress analysis of this device. Finite element analysis nonlinear material capabilities are being used to study, predict and correlate the machine behavior

  1. Bragg rotor spectrometer for tokamak diagnostics

    International Nuclear Information System (INIS)

    Barnsley, R.; Evans, K.D.; Peacock, N.J.

    1986-01-01

    This paper discusses high thru'put broad band (1-24 angstrom) x-ray spectrometer having absolute calibration for wavelength and intensity and demonstrated on the DITE tokamak. This instrument has a self-contained vacuum system which allows full spatial scans of the DITE plasma. Data acquisition and drive mechanism for the rotor and filter selection are operated remotely from a SADA. Results are presented of fast spectral surveys and λ-lock time-evolution of impurity emission during neutral beam injection. Spatial scans of the absolute impurity concentrations are derived

  2. Turbulence near the wall in tokamaks

    International Nuclear Information System (INIS)

    Nedospasov, A.V.

    1989-01-01

    Causes of plasma turbulence in the shadow of poloidal limiter of tokamak are discussed. Current to limiter along the magnetic field plays the determining role. In this case the plasma on the external surrounding of torr is unstable relatively to channeled perturbations. Conditions of strong and weak recycling in limiter are considered. It is shown that in case of strong recycling, the amplitude and scale of turbulence pulsations and turbulent temperature conductivity of the plasma can be determined by low temperature of electrons near the limiter surface

  3. Electrical characteristics of an ideal tokamak limiter

    International Nuclear Information System (INIS)

    Motley, R.W.

    1981-01-01

    The intrusion of an equipotential poloidal limiter into the edge plasma of a circular tokamak discharge distorts the axisymmetry in several ways: (1) it (partially) shorts out the top-to-bottom Pfirsch-Schlueter potentials, (2) it creates zones of reversed equilibrium current flow into the limiter, and (3) it generates an electrostatic field opposing the loop current. The resulting boundary mismatch between the outer layers and the inner axisymmetric Pfirsch-Schlueter layer increases the free energy available to drive the edge plasma unstable. A number of special limiters are proposed to symmetrize the edge plasma and thereby reduce the electrical and MHD activity in the boundary layer. (author)

  4. 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

  5. Time evolution of tokamak states with flow

    International Nuclear Information System (INIS)

    Kerner, W.; Weitzner, H.

    1985-12-01

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

  6. Alpha transport and blistering in tokamaks

    International Nuclear Information System (INIS)

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

    1978-12-01

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

  7. Diamagnetic flux measurement in Aditya tokamak

    International Nuclear Information System (INIS)

    Kumar, Sameer; Jha, Ratneshwar; Lal, Praveen; Hansaliya, Chandresh; Gopalkrishna, M. V.; Kulkarni, Sanjay; Mishra, Kishore

    2010-01-01

    Measurements of diamagnetic flux in Aditya tokamak for different discharge conditions are reported for the first time. The measured diamagnetic flux in a typical discharge is less than 0.6 mWb and therefore it has required careful compensation for various kinds of pick-ups. The hardware and software compensations employed in this measurement are described. We introduce compensation of a pick-up due to plasma current of less than 20 kA in short duration discharges, in which plasma pressure gradient is supposed to be negligible. The flux measurement during radio frequency heating is also presented in order to validate compensation.

  8. Differential and Integral Models of TOKAMAK

    Directory of Open Access Journals (Sweden)

    Ivo Dolezel

    2004-01-01

    Full Text Available Modeling of 3D electromagnetic phenomena in TOKAMAK with typically distributed main and additional coils is not an easy business. Evaluated must be not only distribution of the magnetic field, but also forces acting in particular coils. Use of differential methods (such as FDM or FEM for this purpose may be complicated because of geometrical incommensurability of particular subregions in the investigated area or problems with the boundary conditions. That is why integral formulation of the problem may sometimes be an advantages. The theoretical analysis is illustrated on an example processed by both methods, whose results are compared and discussed.

  9. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1998-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  10. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  12. Saturation of alpha particle driven instability in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gorelenkov, N.N.; Chen, Y.; White, R.B.; Berk, H.L.

    1999-01-01

    A nonlinear theory of kinetic instabilities near threshold [Berk et al., Plasma Phys. Rep. 23, 842 (1997)] is applied to calculate the saturation level of toroidicity-induced Alfven eigenmodes (TAE), and to be compared with the predictions of δf method calculations (Y. Chen, Ph.D. thesis, Princeton University, 1998). Good agreement is observed between the predictions of both methods and the predicted saturation levels are comparable to experimentally measured amplitudes of the TAE oscillations in Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)]. copyright 1999 American Institute of Physics

  13. Magnetic field structure of experimental high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Deniz, A.V.

    1986-01-01

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

  14. Dynamic stabilization of disruption precursors in tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Maoquan, Wang; Jianshan, Mao; Yuan, Pan [Academia Sinica, Hefei, AH (China). Inst. of Plasma Physics

    1994-12-01

    A method for dynamic stabilization of the disruption precursors in tokamak is proposed, that is a controlled ac current induced and added to the equilibrium current. The ac currents applied can be a sine alternative current with a relevant frequency, or a pulsed current with a suitable pulsed width {tau} and or a discontinuous pulsed current whose width {tau} is very shorter than the intervals between pulses, and or a `sawtooth` pulsed current with the time of ramp phase of the sawtooth is very much shorter than the sawtooth descending time, the ratio of them can be {<=}10{sup -3}. The physical model of the ac current drive is analyzed in detail. The suppression role of the ac current on the MHD perturbations was analyzed in theory and proved numerically. It is indicated that the ac current can make the discontinuous derivative, {Delta}`, more favorable for the tearing mode stabilities, and so, as long as the parameters of the applied ac currents are selected suitably, the MHD perturbations can be suppressed effectively, the perturbations will be in the zero-growing state, the profile of the plasma current and temperature remain in the initial states and not variate basically, the tokamak be in the stabilized operation state. (8 figs.).

  15. The stability margin on EAST tokamak

    International Nuclear Information System (INIS)

    Jin-Ping, Qian; Bao-Nian, Wan; Biao, Shen; Bing-Jia, Xiao; Walker, M.L.; Humphreys, D.A.

    2009-01-01

    The experimental advanced superconducting tokamak (EAST) is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma due to the necessary thermal isolation from the superconducting magnets, which leads to relatively weaker coupling between plasma and poloidal field. This may cause more difficulties in controlling the vertical instability by using the poloidal coils. The measured growth rates of vertical stability are compared with theoretical calculations, based on a rigid plasma model. Poloidal beta and internal inductance are varied to investigate their effects on the stability margin by changing the values of parameters α n and γ n (Howl et al 1992 Phys. Fluids B 4 1724), with plasma shape fixed to be a configuration with k = 1.9 and δ = 0.5. A number of ways of studying the stability margin are investigated. Among them, changing the values of parameters κ and l i is shown to be the most effective way to increase the stability margin. Finally, a guideline of stability margin M s (κ, l i , A) to a new discharge scenario showing whether plasmas can be stabilized is also presented in this paper

  16. ICRF heating experiments in JFT-2 tokamak

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1986-01-01

    This is an experimental study of ICRF heating on JFT-2 Tokamak in Japan Atomic Energy Research Institute. In this study, we first clarified physical and engineering problems of ICRF heating of tokamak plasma. Next, we optimized the design of the ICRF heating system, and the plasma parameters for the heating. Finally, we could demonstrate a high efficiency of this additional heating method by launching RF power which is two or three times as large as an ohmic input power to a plasma. And we achieved following things. (1) We optimized a design of an antenna, and we improved a durability of the system for high voltage. With the result that we achieved the maximum power density on an antenna. (2) We demonstrated that electron heating regime and ion heating regime can be easily accessed by controlling plasma parameters. Also we found the optimum heating conditions in each heating regime. (3) We experimentally clarified the production mechanism of impurities during ICRF heating. We could reduce the influx of metal impurity ions to a plasma by employing low z materials for limiters and antenna shields. Consequently, we improved a heating efficiency of electrons. Next, we studied a power balance of plasma during ICRF heating, and we could compare heating characteristics of ICRF with other additional heatings on JFT-2. (author)

  17. Theory of tokamak resistive fishbone modes

    International Nuclear Information System (INIS)

    Shi Bingren; Sui Guofang

    1995-12-01

    A special kind of internal kink mode, the fishbone, can be excited by the energetic particles in tokamak plasmas. Theoretical analyses of fishbone modes based on the ideal MHD framework have predicted that two branches of modes exists. One is the Chen-White branch with ω∼ω-bar dm , corresponding to a higher threshold in β h ; the other is the Coppis branch with ω∼ω *i , and a much lower threshold in β h . The latter mode would put a rather unfavourable restriction on heating efficiency and on plasma confinement. However. It is found that the resistivity effect is essential for this mode. In this paper, a new resistive fishbone mode analysis is carried out. In the (γ mhd ,β H ) space, the stability diagram shows complicate structure, the Coppis branch is replaced by a weakly unstable mode and there is no longer closed stable region. The growth rate of this mode varies with β h , its peak value is still very low compared to other internal modes. The implications of these results to future tokamak experiments are discussed. (8 figs.)

  18. Plasma transport in a compact ignition tokamak

    International Nuclear Information System (INIS)

    Singer, C.E.; Ku, L.P; Bateman, G.

    1987-02-01

    Nominal predicted plasma conditions in a compact ignition tokamak are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models which have given almost equally good fits to experimental data. Using a transport model which best fits the data, thermonuclear ignition occurs in a Compact Ignition Tokamak design with major radius 1.32 m, plasma half-width 0.43 m, elongation 2.0, and toroidal field and plasma current ramped in six seconds from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the 3 He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates, have a large effect on ignition and on the maximum beta that can be achieved

  19. Alfven wave studies on a tokamak

    International Nuclear Information System (INIS)

    Kortbawi, D.

    1987-10-01

    The continuum modes of the shear Alfven resonance are studied on the Tokapole II device, a small tokamak operated in a four node poloidal divertor configuration. A variety of antenna designs and the efficiency with which they deliver energy to the resonant layer are discussed. The spatial structure of the driven waves is studied by means of magnetic probes inserted into the current channel. In an attempt to optimize the coupling of energy in to the resonant layer, the angle of antenna currents with respect to the equilibrium field, antenna size, and plasma-to-antenna distance are varied. The usefulness of Faraday shields, particle shields, and local limiters are investigated. Antennas should be well shielded, either a dense Faraday shield or particle shield being satisfactory. The antenna should be large and very near to the plasma. The wave magnetic fields measured show a spatial resonance, the position of which varies with the value of the equilibrium field and mass density. They are polarized perpendicular to the equilibrium field. A wave propagates radially in to the resonant surface where it is converted to the shear Alfven wave. The signal has a short risetime and does not propagate far toroidally. These points are all consistent with a strongly damped shear Alfven wave. Comparisons of this work to theoretical predictions and results from other tokamaks are made

  20. Mechanism for rapid sawtooth crashes in tokamaks

    International Nuclear Information System (INIS)

    Aydemir, A.Y.; Hazeltine, R.D.

    1986-09-01

    The sawtooth oscillations in the soft x-ray signals observed in tokamaks are associated with periodic changes in the central electron temperature, T/sub e/. Typically, a slow phase during which the central temperature slowly rises is followed by a fast drop in T/sub e/, associated with flattening of the central temperature. The time scale of the slow phase is determined by various transport processes such as ohmic heating. The resistive internal kink mode was invoked by Kadomtsev to explain the crash phase of the oscillations. Fast crash times observed in the large tokamaks are studied here, especially the fast crashes observed in JET. These sawtooth oscillations are characterized by the absence of any discrenible precursor oscillations, and a rapid collapse of the central temperature in about 100 microseconds. During the crash phase, the hot core region rapidly moves outward and is replaced by colder plasma. Then, this highly asymmetric state relaxes (in ∼100μsec) to a poloidally symmetric state in which a ring of hot plasma surrounds the colder core plasma, producing a hollow pressure profile

  1. Electron cyclotron heating (ECH) of tokamak plasmas

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi

    1990-01-01

    Electron cyclotron heating (ECH) is one of the intense methods of plasma heating, and which utilizes the collisionless electron-cyclotron-resonance-interaction between the launched electromagnetic waves (called electron cyclotron waves) and electrons which are one of the constituents of the high temperature plasmas. Another constituent, namely the ions which are subject to nuclear fusion, are heated indirectly but strongly and instantly (in about 0.1 s) by the collisions with the ECH-heated electrons in the fusion plasmas. The recent progress on the development of high-power and high-frequency millimeter-wave-source enabled the ECH experiments in the middle size tokamaks such as JFT-2M (Japan), Doublet III (USA), T-10 (USSR) etc., and ECH has been demonstrated to be the sure and intense plasma heating method. The ECH attracts much attention for its remarkable capabilities; to produce plasmas (pre-ionization), to heat plasmas, to drive plasma current for the plasma confinement, and recently especially by the localization and the spatial controllability of its heating zone, which is beneficial for the fine controls of the profiles of plasma parameters (temperature, current density etc.), for the control of the magnetohydrodynamic instabilities, or for the optimization/improvement of the plasma confinement characteristics. Here, the present status of the ECH studies on tokamak plasmas are reviewed. (author)

  2. Overview of the Tokamak de Varennes program

    International Nuclear Information System (INIS)

    Pacher, H.D.

    1986-01-01

    The Tokamak de Varennes will be the major Canadian experiment in the magnetic fusion domain. It has a toroidal field of 1.5 tesla, major radius of 0.85 m, a minor radius of 0.25 m, and will study long pulses, up to 30 seconds duration. Initially, a series of successive plasma pulses, each of the order of seconds, will yield a duty factor of over 50 percent. During this phase, the major emphasis will be on the study of impurity generation, transport, and control, plasma-wall interactions and material properties. The program will include studies of fast current rampdown and the resultant current profile modifications. The development of advanced diagnostics will also be undertaken. To attain a higher duty factor with continuous plasma operation, noninductive current drive by radio=frequency will be added as an early upgrade. This will introduce current drive investigations such as transformer recharge and profile relaxation, and enhance the wall and materials study program. In this context, the Tokamak de Varennes will concentrate on the study of impurity exhaust and retention as well as net erosion of the limiter and neutralization plate materials

  3. Fractional power operation of tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Vold, E.L.; Conn, R.W.

    1986-01-01

    Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

  4. Simulation of saturated tearing modes in tokamaks

    International Nuclear Information System (INIS)

    Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.

    2004-01-01

    A quasi-linear model, which includes the effect of the neoclassical bootstrap current, is developed for saturated tearing modes in order to compute magnetic island widths in axisymmetric toroidal plasmas with arbitrary aspect ratio and cross-sectional shape. The model is tested in a simple stand-alone code and is implemented in the BALDUR [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1982)] predictive modeling code. It is found that the widths of tearing mode islands increase with decreasing aspect ratio and with increasing elongation. Also, the island widths increase when the gradient of the current density increases at the edge of the islands and when the current density inside the islands is suppressed, such as the suppression caused by the near absence of the bootstrap current within the islands. In simulations of tokamak discharges, it is found that tearing mode island widths oscillate in time in response to periodic sawtooth crashes. The local enhancements in the transport produced by magnetic islands have a noticeable effect on global plasma confinement in simulations of low aspect ratio, high beta tokamaks, where saturated tearing mode islands can occur with widths that are greater than 15% of the plasma minor radius

  5. Anomalous transport in the tokamak edge

    International Nuclear Information System (INIS)

    Vayakis, G.

    1991-04-01

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

  6. Simulation of tokamak runaway-electron events

    International Nuclear Information System (INIS)

    Bolt, H.; Miyahara, A.; Miyake, M.; Yamamoto, T.

    1987-08-01

    High energy runaway-electron events which can occur in tokamaks when the plasma hits the first wall are a critical issue for the materials selection of future devices. Runaway-electron events are simulated with an electron linear accelerator to better understand the observed runaway-electron damage to tokamak first wall materials and to consider the runaway-electron issue in further materials development and selection. The electron linear accelerator produces beam energies of 20 to 30 MeV at an integrated power input of up to 1.3 kW. Graphite, SiC + 2 % AlN, stainless steel, molybdenum and tungsten have been tested as bulk materials. To test the reliability of actively cooled systems under runaway-electron impact layer systems of graphite fixed to metal substrates have been tested. The irradiation resulted in damage to the metal compounds but left graphite and SiC + 2 % AlN without damage. Metal substrates of graphite - metal systems for actively cooled structures suffer severe damage unless thick graphite shielding is provided. (author)

  7. A tokamak reactor with servicing capability

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  8. Modular pump limiter systems for large tokamaks

    International Nuclear Information System (INIS)

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

    1987-09-01

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

  9. Theoretical scaling law for ohmically heated tokamaks

    International Nuclear Information System (INIS)

    Minardi, E.

    1981-06-01

    The electrostatic drift instability arising from the reduction of shear damping, due to toroidal effects, is assumed to be the basic source of the anomalous electron transport in tokamaks. The Maxwellian population of electrons constitutes a medium whose adiabatic nonlinear reaction to the instability (described in terms of an effective dielectric constant of the medium) determines the stationary electrostatic fluctuation level in marginally unstable situations. The existence of a random electrostatic potenial implies a fluctuating current of the Maxwellian electrons which creates a random magnetic field and a stocasticization of a magnetic configuration. The application of recent results allows the calculation of the realted radial electron transport. It is found that the confinement time under stationary ohmic conditions scales as n Tsub(i)sup( - 1/2) and is proportional roughly to the cube of the geometric dimenisions. Moreover, it is deduced that the loop voltage is approximateley the same for all tokamaks, irrespective of temperature and density and to a large extent, also of geometrical conditions. Thes results are characteristic of the ohmic stationary regime and can hardly be extrapolated to order heating regimes. (orig.)

  10. MDSplus integration at TCABR tokamak: Current status

    International Nuclear Information System (INIS)

    Sá, W.P. de; Ronchi, G.

    2016-01-01

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  11. Integrated plasma control for high performance tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.

    2005-01-01

    Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)

  12. System studies of compact ignition tokamaks

    International Nuclear Information System (INIS)

    Galambos, J.D.; Peng, Y.K.M.; Blackfield, D.T.

    1986-01-01

    A new version of the FEDC Tokamak System Code (TSC) has been developed to analyze the Compact Ignition Tokamak (CIT). These proposed experiments have small (major radius F 1.5m) and high magnetic fields (B J 10T), and are characterized by reduced cost. Key design constraints of CIT include limits to the high stress levels in the magnetic coils, limits to the large temperature rises in the coils and on the first wall or divertor plate, minimizing power supply requirements, and assuring adequate plasma performance in fusion ignition and burn time consistent with the latest physics understanding. We present systems code level studies of CIT parameter space here for a range of design options with various design constraints. The present version of the TSC incorporates new models for key components of CIT. For example, new algorithms have been incorporated for calculating stress levels in the TFC and ohmic solenoid, temperature rise in the magnetic coils, peak power requirements, plasma MHD equilibrium and volt-second capability. The code also incorporates a numerical optimizer to find combinations of engineering quantities (device size, coil sizes, coil current densities etc.) and physics quantities (plasma density temperature, and beta, etc.) which satisfy all the constraints and can minimize or maximize a figure of merit (e.g., the major radius). This method was recently used in a mirror reactor system code (3) for the Minimara concept development

  13. Experimental investigations at the Soviet tokamaks

    International Nuclear Information System (INIS)

    Bobrovskij, G.A.; Golant, V.E.; AN SSSR, Leningrad. Fiziko-Tekhnicheskij Inst.)

    1978-01-01

    The review is devoted to the basic results obtained on the Soviet tokamaks during 1976-1977. Behaviour of impurities, tearing instability, additional methods of plasma heating, energy distribution function were investigated. A brief description of new T-7, TM-4, ''Tuman-3'' tokamaks is given. It is shown that despite inflow of impurities to the pinch periphery, no their appreciable accumulation is observed at least during the discharge time. It is shown that the helical perturbations with m=2 and 1 present the greatest danger. The suppression of the tearing instability is related with suppression of the mode with m=2. The helical perturbation prevents formation of skin configuration at the initial stage of the discharge. As a rule, the transition of an appreciable fraction of electrons to continuous acceleration does not take place, although a significant deformation of electron distribution function under the action of electric field occurs. Plasma compression by increasing magnetic field induces oscillations and improves thermal plasma isolation. It is shown experimentally that the considerable efficiency of energy contribution to the ion component at the central part of plasma may be obtained by means of HF heating under conditions of low-hybrid resonance. It is shown that the recombination has a considerable effect on concentration of neutral particles in the central region

  14. Physics evaluation of compact tokamak ignition experiments

    International Nuclear Information System (INIS)

    Uckan, N.A.; Houlberg, W.A.; Sheffield, J.

    1985-01-01

    At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/ 2 /q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs

  15. MDSplus integration at TCABR tokamak: Current status

    Energy Technology Data Exchange (ETDEWEB)

    Sá, W.P. de, E-mail: pires@if.usp.br; Ronchi, G., E-mail: gronchi@if.usp.br

    2016-11-15

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  16. Ion cyclotron system design for KSTAR tokamak

    International Nuclear Information System (INIS)

    Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H.

    1998-05-01

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs

  17. Ion cyclotron system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H

    1998-05-01

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs.

  18. 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.

  19. 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

  20. Tokamak advanced pump limiter experiments and analysis

    International Nuclear Information System (INIS)

    Conn, R.W.

    1983-06-01

    Experiments with pump limiter modules on several operating tokamaks establish such limiters as efficient collectors of particles and has demonstrated the importance of ballistic scattering as predicted theoretically. Plasma interaction with recycling neutral gas appears to become important as the plasma density increases and the effective ionization mean free path within the module decreases. In limiters with particle collection but without active internal pumping, the neutral gas pressure is found to vary nonlinearly with the edge plasma density at the highest densities studies. Both experiments and theory indicate that the energy spectrum of gas atoms in the pump ducting is non-thermal, consistent with the results of Monte Carlo neutral atom transport calculations. The distribution of plasma power over the front surface of such modules has been measured and appears to be consistent with the predictions of simple theory. Initial results from the latest experiment on the ISX-B tokamak with an actively pumped limiter module demonstrates that the core plasma density can be controlled with a pump limiter and that the scrape-off layer plasma can partially screen the core plasma from gas injection. The results from module pump limiter experiments and from the theory and design analysis of advanced pump limiters for reactors are used to suggest the major features of a definitive, axisymmetric, toroidal belt pump limiter experiment

  1. Counterstreaming-ion-tokamak fissile breeder

    International Nuclear Information System (INIS)

    Jassby, D.L.; Lee, J.D.

    1976-08-01

    Tokamak plasmas fueled and heated by energetic neutral-atom beams are characterized by total ion energy greatly exceeding the electron energy. For smaller devices the largest fusion reactivity of energetic-ion plasmas is obtained when oppositely injected D 0 and T 0 beams sustain counterstreaming velocity distributions of deuterons and tritons. This scoping study investigates the net fissile and power productions of a tokamak fusion-fission reactor with a counterstreaming-ion fusion driver and a fertile blanket optimized for fissile breeding. The fusion driver has parameters R/sub o/ = 4.7 m, a = 1.0 m, B/sub t/ = 5.6 T, W/sub b/ = 100 keV (D 0 ), n tau/sub E/ = 1.4 x 10 13 cm -3 s, Q = 1.5, 14-MeV neutron production = 175 MW. The blanket contains a fast-fission zone of natural U plus Mo (7 percent), followed by a Li-bearing zone for T breeding. The reactor produces a net power of 480 MWe and supplies sufficient Pu to support a system of LWR's producing 3800 MWe, with an estimated electrical energy cost for the entire system of 27 mills/kWh

  2. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Nicolas, T.

    2013-01-01

    The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr

  3. Thomson scattering on the PRETEXT Tokamak

    International Nuclear Information System (INIS)

    McCool, S.C.

    1982-03-01

    Ruby laser Thomson scattering was performed on the PRETEXT tokamak. A 10 Joule Q-switched laser and a 1 meter 10 channel polychromator were used to diagnose the electron temperature and density profiles in the PRETEXT plasma. These parameters were measured as a function of time and radial position on a shot to shot basis. The density measurement was calibrated by Rayleigh and Raman scattering and by comparison with data from a 4 mm microwave interferometer. Electron densities ranging from 1 x 10 12 cm -3 to 2 x 10 13 cm -3 and temperatures ranging from 3 eV to 400 eV were observed. Detailed measurements were made throughout the 40 ms discharge with particular emphasis on the current rise phase. The Thomson scattering data was used as input to a one dimensional magnetic diffusion code. This code modelled the evolution of the current density and safety factor profiles. The results of this analysis were compared with existing theories of tokamak current penetration. The growth of resitive MHD tearing modes was proposed as a likely explanation for the anomalously rapid current penetration observed in PRETEXT

  4. Tokamak power systems studies at ANL

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Neutron measurement techniques for tokamak plasmas

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1994-01-01

    The present article reviews the neutron measurement techniques that are currently being applied to the study of tokamak plasmas. The range of neutron energies of primary interest is limited to narrow bands around 2.5 and 14 MeV, and the variety of measurements that can be made for plasma diagnostic purposes is also restricted. To characterize the plasma as a neutron source, it is necessary only to measure the total neutron emission, the relative neutron emissivity as a function of position throughout the plasma, and the energy spectra of the emitted neutrons. In principle, such measurements might be expected to be relatively easy. That this is not the case is, in part, attributable to practical problems of accessibility to a harsh environment but is mostly a consequence of the time-scale on which the measurements have to be made and of the wide range of neutron emission intensities that have to be covered: for tokamak studies, the time-scale is of the order of 1 to 100 ms and the neutron intensity ranges from 10 12 to 10 19 s -1 . (author)

  6. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  7. 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)

  8. 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)

  9. 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.

  10. Structured Cable for High-Current Coils of Tokamaks

    Science.gov (United States)

    Benson, Christopher; McIntyre, Peter; Sattarov, Akhdiyor; Mann, Thomas

    2011-10-01

    The 45 kA superconducting cable for the ITER central solenoid coil has yielded questionable results in two recent tests. In both cases the cable Tc increased after cycling only a fraction of the design life, indicating degradation due to fatigue and fracture among the superconducting strands. The Accelerator Research Lab at Texas A&M University is developing a design for a Nb3Sn structured cable suitable for such tokamak coils. The superconductor is configured in 6 sub-cables, and each subcable is supported within a channel of a central support structure within a high-strength armor sheath. The structured cable addresses two issues that are thought to compromise opposition at high current. The strands are supported without cross-overs (which produce stress concentration); and armor sheath and core structure bypass stress through the coil and among subcables so that the stress within each subcable is only what is produced directly upon it. Details of the design and plans for development will be presented.

  11. Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

    1979-04-01

    The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

  12. Technology and plasma-materials interaction processes of tokamak disruptions

    International Nuclear Information System (INIS)

    McGrath, R.T.; Kellman, A.G.

    1992-01-01

    A workshop on the technology and plasma-materials interaction processes of tokamak disruptions was held April 3, 1992 in Monterey, California, as a satellite meeting of the 10th International Conference on Plasma-Surface Interactions. The objective was to bring together researchers working on disruption measurements in operating tokamaks, those performing disruption simulation experiments using pulsed plasma gun, electron beam and laser systems, and computational physicists attempting to model the evolution and plasma-materials interaction processes of tokamak disruptions. This is a brief report on the workshop. 4 refs

  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. Digital control of plasma position in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

    2002-03-01

    Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.

    1986-10-01

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

  16. Engineering design of TFTR and it's impact on future tokamaks

    International Nuclear Information System (INIS)

    Sabado, M.M.

    1981-01-01

    TFTR is a second generation tokamak whose key objective is scientific break-even. TFTR is expected to be the first machine to demonstrate proper combination of plasma confinement time, density, and temperature to obtain this objective. A summary of major TFTR design parameters, including TFM, is presented, and their potential impact on future tokamaks discussed. Details of the updated engineering design and analysis of components are described. Status of major hardware fabrication, assembly installation and test are reviewed. TFTR features, technology, predicted performance and their potential implication for future tokamaks are summarized

  17. 3He functions in tokamak-pumped laser systems

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-10-01

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

  18. Particle injection into the Castor tokamak by electric arcs

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

    1989-01-01

    The influence of arcing on the tokamak discharge was investigated in the Castor tokamak. A special calibrated gun which emitted tantalum by artificially ignited electric arcs, was used to study the transport of the injected tantalum ions, neutrals and droplets. The injection of tantalum led to an increase in electron density and to a change of plasma position only if the transported charge was higher than 0.01 C. As the naturally occurring arcs are well below this limit, the arcing in tokamaks is rather the consequence than the reason of instabilities. (J.U.)

  19. The residual zonal dynamics in a toroidally rotating tokamak

    International Nuclear Information System (INIS)

    Zhou Deng

    2015-01-01

    Zonal flows, initially driven by ion-temperature-gradient turbulence, may evolve due to the neoclassic polarization in a collisionless tokamak plasma. In this presentation, the form of the residual zonal flow is presented for tokamak plasmas rotating toroidally at arbitrary velocity. The gyro-kinetic equation is analytically solved to give the expression of residual zonal flows with arbitrary rotating velocity. The zonal flow level decreases as the rotating velocity increases. The numerical evaluation is in good agreement with the previous simulation result for high aspect ratio tokamaks. (author)

  20. 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