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Sample records for aditya tokamak research

  1. SST and ADITYA tokamak research in India

    Steady state operation of tokamaks plays an important role in high temperature magnetically confined plasma research. Steady state Superconducting Tokamak (SST) programme in India deals with the development of various technologies in this direction. SST-1 machine has been engineered and is being fabricated at the Institute for Plasma Research. The objectives of the machine are to study physics of plasma processes under steady state condition and develop the technologies related to steady state operation. Various sub-systems are being prototyped and developed. SST-1 is a large aspect ratio machine with a major radius of 1.1 m and a plasma minor radius of 0.2 m with elongation of 1.7 to 1.9 and triangularity of 0.5 to 0.7. It has been designed for 1000 sec operation at 3 T toroidal magnetic eld. Neutral beam Injection and Radio frequency heating systems are being developed to heat the plasma. Lower hybrid Current Drive system would sustain 200 kA of plasma current during 1000 sec operation. ADITYA tokamak has been upgraded with new diagnostics and RF heating systems. Thomson Scattering and ECE diagnostics have been operated. 200 kW Ion Cyclotron Resonance Heating (ICRH) and 200 kW Electron Cyclotron Resonance Heating (ECRH) systems have been successfully commissioned. RF assisted initial breakdown experiments have been initiated with these systems. (author)

  2. The upgradation of Aditya Tokamak

    Aditya Tokamak is the first Indian tokamak, indigenously built and commissioned at the Institute for Plasma Research, Gandhinagar, Gujarat, India, in September, 1989. Aditya Tokamak has been in operation since more than 25 years. More than 30,000 discharges are taken and a large number of experiments are carried out, with plasma current ranging from 50 KA to 150 KA, lasting for 100 to 250 milliseconds. Various types of wall conditioning techniques and different hot plasma diagnostics are tested and operated on Aditya Tokamak. The experiments for turbulent particle transport and turbulence in the edge plasma, gas puffing, lithium coating, mitigation, plasma disruption, limiter and electron biasing, runaway discharges etc. led to many interesting results contributing immensely to the world of thermonuclear fusion. Experiments on Pre-ionization and Plasma heating by ICRH and ECRH are also worked out. The scientific objectives of Aditya tokamak Upgrade include Low loop voltage plasma start-up with strong pre-ionization having a good plasma control system. The upgrade is designed keeping in mind the experiments, disruption mitigation studies relevant to future fusion devices, runway mitigation studies, demonstration of Radio-frequency heating and current drive etc. This upgraded Aditya tokamak will be used for basic studies on plasma confinement and scaling to larger devices, development and testing of new diagnostics etc. This machine will be easily accessible compared to SST-1 and will be very useful for generation of technical and scientific expertise for future fusion devices. In this paper, especial features of the upgrade including various aspects of designing of new components for Aditya Upgrade tokamak is presented

  3. Assembly of Aditya upgrade tokamak

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a tokamak with divertor configuration. At present the existing ADITYA tokamak has been dismantled up to bottom plinth on which the whole assembly of toroidal field (TF) coils and vacuum vessel rested. The major components of ADITYA machine includes 20 TF coils and its structural components, 9 Ohmic coils and its clamps, 4 BV coils and its clamps as well as their busbar connections, vacuum vessel and its supports and buckling cylinder, which are all being dismantled. The re-assembly of the ADITYA Upgrade tokamak started with installation and positioning of new buckling cylinder and central solenoid (TR1) coil. After that the inner sections of TF coils are placed following which in-situ winding, installation, positioning and support mounting of two pairs of new inner divertor coils have been carried out. After securing the TF coils with top I-beams the new torus shaped vacuum vessel with circular cross-section in 2 halves have been installed. The assembly of TF structural components such as top and bottom guiding wedges, driving wedges, top and bottom compression ring, inner and outer fish plates and top inverted triangle has been carried out in an appropriate sequence. The assembly of outer sections of TF coils along with the proper placements of top auxiliary TR and vertical field coils with proper alignment and positioning with the optical metrology instrument mainly completes the reassembly. Detailed re-assembly steps and challenges faced during re-assembly will be discussed in this paper. (author)

  4. Metrology measurements for Aditya tokamak upgradation

    After 25 years of Aditya tokamak (midsized, air-core, R0= 75 cm, a = 25 cm) operation achieving high temperature circular plasmas in limiter configuration, upgrading it to Aditya-U tokamak with divertor configuration has been planned and the upgradation is under progress. The upgradation process include dismantling of the existing Aditya tokamak to its base level and re-erect it by placing new subsystems like new vacuum vessel of circular cross-section, new buckling cylinder etc. Apposite alignment of subsystems, mainly all the magnetic coil systems in all grades and scales of tokamak is very crucial and essential, as misaligned magnetic coil system scan generate error magnetic fields, which can significantly impact the plasma formation and sustainment in a tokamak. With this motivation, position and alignment measurement of the existing magnetic coils and structural components of ADITYA tokamak is carried out for the very first time with the optical metrology instrument. Prior to carrying out measurement exercise, machine datum has been transferred to the reference on the wall of tokamak hall using five-point laser and the machine center has been transformed to the four wall of tokamak hall. All position measurements are done with respect to machine major axis in cylindrical geometry. Measurement includes existing radial (R) and elevation (Z) positions of all magnetic coils and various structural components within the accuracy of ± 1 mm. More than 5000 data points are recorded using optical metrology instrument. Again the elevation references are transferred to the primary network established and the angular references are transformed on the floor of the tokamak hall. These results will serve as ready reference for reassembly and alignment of Aditya - Upgrade tokamak. In this paper detailed position measurements of different subsystems of old Aditya tokamak and the relocation of them along with new ones using the optical metrology instruments will be presented

  5. Recent experiments and upgradation plans for Aditya Tokamak

    Several experiments relevant to the operation of future big tokamaks including ITER and also contributing significantly to the tokamak based thermonuclear fusion research have been carried out in Aditya tokamak recently. Low loop voltage start-up of plasma current has been successfully obtained with ICR and ECR preionization. Reduced runaway generation is achieved by applying a local vertical magnetic field at one toroidal location. Plasma disruptions, a sudden loss of equilibrium and confinement, has been successfully mitigated through application of bias voltage on a Molybdenum (Mo) electrode placed inside the last closed flux surface (LCFS). Extensive studies on plasma flows, effect of gas puff on flows in the Scrape off layer (SOL) and impurity transport has been carried out. Effect of Helium glow discharge cleaning (GDC) on partial pressures and plasma parameters have also been studied for plasma performance improvement. To contribute more to the bigger tokamaks operated in the divertor configuration, the existing Aditya tokamak with limiter configuration, which is in operation for 24 years, is planned to be upgraded to a divertor machine. The main aim of the Aditya-U tokamak is to have a small/mid-size tokamak with divertor operation and higher duty cycle, which will be test bed for new diagnostics, the students can be trained and those experiments can be tried out which are not desirable in big tokamaks, such as runaway mitigation and disruption control. Details of experimental results and upgradation plan will be discussed in the talk. (author)

  6. Radiation power measurement on the ADITYA tokamak

    Tahiliani, Kumudni; Jha, Ratneshwar; Gopalkrishana, M. V.; Doshi, Kalpesh; Rathod, Vipal; Hansalia, Chandresh; ADITYA Team

    2009-08-01

    The radiation power loss and its variation with plasma density and current are studied in the ADITYA tokamak. The radiation power loss varies from 20% to 40% of the input power for different discharges. The radiation fraction decreases with increasing plasma current but it increases with increasing line-averaged central density. The radiated power behavior has also been studied in discharges with short pulses of molecular beam injection (MBI) and gas puff (GP). The increase in radiation loss is limited to the edge chords in the case of GP, but it extends to the core region for MBI fueling. The MBI seems to indicate reduction in the edge recycling. It is observed that during the density limit disruption, the radiated power loss is more in the current quench phase as compared with the thermal quench phase and comes mainly from the plasma edge.

  7. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    D Raju; R Jha; P K Kaw; S K Mattoo; Y C Saxena; Aditya Team

    2000-11-01

    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 the discharge progresses. It is observed that during the current rise phase, current perturbation undergoes transition from = 5 poloidal structure to = 4 and then to = 3. At the time of current termination, = 2 perturbation is observed. It is observed that the mode frequency remains nearly constant (≈10 kHz) when poloidal mode structure changes from = 4 to = 2. This may be either an indication of mode coupling or a consequences of changes in the plasma electron temperature and density scale length.

  8. Ohmic discharges with improved confinement in Tokamak Aditya

    ADITYA (R0 = 75 cm, a = 25 cm), an ohmically heated circular limiter tokamak is regularly being operated to carry out several experiments related to controlled thermonuclear fusion research. In recent experimental schedule, special efforts are made to enhance the plasma parameters to achieve Ohmic discharges with improved confinement. Repeatable plasma discharges of maximum plasma current of ∼ 160 kA and discharge duration beyond ∼ 250 ms with plasma current flattop duration of ∼ 140 ms has been obtained for the first time in the first Indian tokamak ADITYA. The discharge reproducibility has been improved with Lithium wall conditioning and much-improved plasma discharges are obtained by precisely controlling the plasma position. Improved discharges are attempted over a wider parameter range to carry out various confinement scaling experiments. In these discharges, chord-averaged electron density 1.0 - 4.0 X 1019m-3 using multiple hydrogen gas puffs, plasma temperature of the order of ∼ 400 - 700 eV has been achieved. The measured confinement time matches quite well with ALCATOR scaling for most of the discharges. It is also observed that in new discharges, the confinement time crosses the L-mode scaling. Detailed analysis of these discharges along with the possible reasons for obtaining higher confinement times will be addressed in this paper. (author)

  9. Investigation of Aditya Tokamak plasmas with lithiumized wall

    The lithium coating on plasma facing components of tokamak leads to better plasma properties through the reduction in impurities and controlling the hydrogen recycling. In Aditya tokamak, lithiumization of vacuum vessel wall is regularly carried out prior to its daily operation using lithium rod exposed to overnight glow discharge-cleaning plasma. Spectroscopic studies of Aditya tokamak plasmas shows the reduction of hydrogen (Hα at 656.3 nm) and oxygen (O II at 441.6 nm) as compared to discharges without the lithium coated walls. This clearly indicates reduction of recycling and impurity influxes from the wall, respectively. After Li coating, plasma stored energy increases significantly and plasmas with higher electron densities are obtained. Estimation of energy confinement time shows that it increases after lithimization and becomes comparable to the values predicated by Neo-Alcator scaling for ohmically heated tokamak plasma. Further analysis indicates that recycling must be low to achieve better plasma confinement. (author)

  10. Conceptual design of PAM antenna for Aditya-U Tokamak

    ADITYA Tokamak is being upgraded (ADITYA-U) to operate the machine at enhanced plasma parameter. This also provides an opportunity to upgrade lower hybrid current drive (LHCD) system to drive plasma current non-inductively and enhance the coupling of RF power to the plasma. It is proposed to replace existing grill antenna by a new type of antenna which is often referred as passive active multijunction (PAM) antenna. The PAM antenna has an advantage of providing efficient RF coupling to the plasma, even at edge densities close to cut-off. Further it provides a lower reflection from the plasma as compared to the conventional grill antenna. ADITYA-U would operate at toroidal magnetic field of 1.5T and may have line average density lying in the range of (0.8 - 3.0) X 1019 m-3. For LHW's to access to the plasma center, the waves would be launched having parallel refractive index (N∥) which is well above the critical accessible condition given by Stix. Thus the PAM antenna is designed to launch N∥ of 2.25 ± 0.28. Our analysis reveals that periodicity for the PAM antenna would be 27mm to launch the design value of N∥ with three passive and three active waveguide in a single PAM module having phase shift of 270° between adjacent active waveguides. The size of the radial port (490 mm x 190 mm) of ADITYA-U tokamak determines the number of PAM modules which may be accommodated in the new scheme. It turns out that two modules of PAM antenna may be accommodated in the said radial port. Mode convertors (TE10 to TE30 mode) would be employed for dividing the RF power in three poloidal locations. A thermal and electro-mechanical analysis is also discussed in this paper. (author)

  11. Studies of impurity behavior during lithiumization experiment in Aditya Tokamak

    Coating of plasma facing components mainly the vacuum vessel wall in tokamaks using low Z material is well known for improving the plasma performance in terms of achieving higher temperatures and low impurities. Among various materials used for coating, lithium has become immensely useful to reduce wall recycling and to decrease the plasma impurity content. In Aditya tokamak Lithiumization, carried out by inserting two Lithium rods inside the glow discharge cleaning plasma, is regularly done to study its effect on plasma performance. Impurity behaviors in the plasma after Li coating have been studied using spectroscopic diagnostics containing optical fibers, interference filters, PMT based filter-scopes and a 0.5 m visible spectrometer through the observations of visible spectra from different species. The temporal behavior of emissions from the plasma shows a decrease in Hα emission after lithiumization indicating reduction in wall recycling. Reduction of O II spectral emission intensity at 441.5 nm and visible continuum at 536.0 nm indicates lower oxygen content in plasma and reduced effective charge, respectively. However, no change is observed in CIII signal monitored at 464.7 nm which might be related to its source i.e. carbon graphite Limiter, on which Lithium coating wiped out quickly due its more direct interaction with plasma compared to the vacuum vessel wall. From the behavior of spectral line of neutral lithium at 670.8 nm monitored by spectrometer, it has been found that the lithium coating, obtained by inserting lithium rods in glow discharge plasmas in Aditya tokamak for 12 hours, sustains up to 12 - 14 long (∼ 100 ms) discharges and then gradually fades away. The sputtering yield of lithium has been estimated spectroscopically, which provides many useful information about the plasma wall interaction in Aditya tokamak. (author)

  12. Measurement of LHCD antenna position in Aditya tokamak

    To drive plasma current non-inductively in ADITYA tokamak, 120 kW pulsed Lower Hybrid Current Drive (LHCD) system at 3.7 GHz has been designed, fabricated and installed on ADITYA tokamak. In this system, the antenna consists of a grill structure, having two rows, each row comprising of four sub-waveguides. The coupling of LHCD power to the plasma strongly depends on the plasma density near the mouth of grill antenna. Thus the grill antenna has to be precisely positioned for efficient coupling. The movement of mechanical bellow, which contracts or expands up to 50mm, governs the movement of antenna. In order to monitor the position of the antenna precisely, the reference position of the antenna with respect to the machine/plasma position has to be accurately determined. Further a mechanical system or an electronic system to measure the relative movement of the antenna with respect to the reference position is also desired. Also due to poor accessibility inside the ADITYA machine, it is impossible to measure physically the reference position of the grill antenna with respect to machine wall, taken as reference position and hence an alternative method has to be adopted to establish these measurements reliably. In this paper we report the design and development of a mechanism, using which the antenna position measurements are made. It also describes a unique method employing which the measurements of the reference position of the antenna with respect to the inner edge of the tokamak wall is carried out, which otherwise was impossible due to poor accessibility and physical constraints. The position of the antenna is monitored using an electronic scale, which is developed and installed on the bellow. Once the reference position is derived, the linear potentiometer, attached to the bellow, measures the linear distance using position transmitter. The accuracy of measurement obtained in our setup is within +/- 0.5 % and the linearity, along with repeatability is excellent.

  13. Novel approaches for mitigating runaway electrons and plasma disruptions in ADITYA tokamak

    Tanna, R. L.; Ghosh, J.; Chattopadhyay, P. K.; Dhyani, Pravesh; Purohit, Shishir; Joisa, S.; Rao, C. V. S.; Panchal, V. K.; Raju, D.; Jadeja, K. A.; Bhatt, S. B.; Gupta, C. N.; Chavda, Chhaya; Kulkarni, S. V.; Shukla, B. K.; Praveenlal E., V.; Raval, Jayesh; Amardas, A.; Atrey, P. K.; Dhobi, U.; Manchanda, R.; Ramaiya, N.; Patel, N.; Chowdhuri, M. B.; Jha, S. K.; Jha, R.; Sen, A.; Saxena, Y. C.; Bora, D.; the ADITYA Team

    2015-06-01

    This paper summarizes the results of recent dedicated experiments on disruption control and runaway mitigation carried out in ADITYA, which are of the utmost importance for the successful operation of large size tokamaks, such as ITER. It is quite a well-known fact that disruptions in tokamaks must be avoided. Disruptions, induced by hydrogen gas puffing, are successfully avoided by two innovative techniques in ADITYA using a bias electrode placed inside the last closed flux surface and applying an ion cyclotron resonance pulse with a power of ∼50 to 70 kW. These experiments led to better understanding of the disruption avoidance mechanisms and also can be thought of as one of the options for disruption avoidance in ITER. In both cases, the physical mechanism seems to be the control of magnetohydrodynamic modes due to increased poloidal rotation of edge plasma generated by induced radial electric fields. Real time avoidance of disruption with identifying proper precursors in both the mechanisms is successfully attempted. Further, analysing thoroughly the huge database of different types of spontaneous and deliberately-triggered disruptions from ADITYA, a significant contribution has been made to the international disruption database (ITPA). Furthermore, the mitigation of the runaway electron generated mainly during disruptions remains a challenging topic in present tokamak research as these high-energy electrons can cause severe damage to in-vessel components and the vacuum vessel. A simple technique has been implemented in ADITYA to mitigate the runaway electrons before they can gain high energies using a localized vertical magnetic field perturbation applied at one toroidal location to extract runaway electrons.

  14. Study of neutral particle transport in Aditya Tokamak plasma using DEGAS2 Code

    Aditya tokamak is a medium sized air-core tokamak having a limiter configuration. The circular poloidal ring limiter is placed at one particular toroidal location. The spatial profile of neutral particles are experimentally observed in this tokamak and the observation suggests important roles of charge exchange processes into the penetration of neutral particle in plasma core. Therefore, to understand the neutral dynamics in Aditya tokamak, the neutral particle transport studies have been carried out using the DEGAS2 code. This code is based on Monte Carlo algorithms and extensively used for investigating the dynamics of neutrals in various tokamaks having divertors as the plasma facing component. The required modification has been carried out in the machine geometries and plasma parameter files through the user developed programs for ADITYA tokamak plasma. Modifications are successfully implemented in this code and the radial profile of Hα emissivity has been obtained. The simulated results are then compared with the experimental observations. In this paper, details on the implementation of the code on Aditya tokamak plasmas are presented and the simulation results are compared with the experiments to understand the neutral particle behaviour in Aditya tokamak plasma. (author)

  15. Structural analysis of new vacuum vessel for Aditya Tokamak upgrade

    The new toroidal-shaped vacuum vessel for Aditya Tokamak Upgrade is fabricated by joining two semi tori of circular cross section, equipped with as many as 115 ports of different sizes and shapes for pumping and diagnostics. Both semi tori are identical and are made up of stainless steel 304L. The major radius of toroidal chamber is 750 mm and minor radius is 305 mm. The vacuum vessel is subjected to different loads such as vacuum load and electromagnetic loads. As the vacuum level required inside the vessel is ∼ 1 x 10-9 mbar, the vessel wall should sustain compressive forces due to atmospheric pressure from outside and should not deform. Hence, the wall thickness of the vessel wall has been chosen after carrying out the detailed stress analysis in ANSYS workbench. Meshing has been carried out using the method of Tetrahedron in the workbench. The maximum stress on vessel due to pressure difference comes out to be ∼ 70 MPa. The maximum deformation for a wall thickness of 10 mm is ∼ 0.45 mm. The vacuum vessel is also planned to be baked up to 150 °C, and the maximum stress on vessel due to combined thermal load and vacuum load (10-9 mbar) becomes ∼ 80 MPa and maximum deformation is 2.95 mm for 10 mm thick walls. As the yield strength of stainless steel 304L is 170 MPa, the stress generated by various load acting on vacuum vessel is under safety limit. Detailed design consideration thoroughly substantiated by ANSYS analysis for the new vacuum vessel of Aditya Tokamak Upgrade will be presented in this paper. (author)

  16. Plasma diagnostics at Aditya Tokamak by two views visible light tomography

    Goswami, Mayank, E-mail: mggm1982@gmail.com [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur (India); Munshi, Prabhat [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur (India); Department of Mechanical Engineering, Indian Institute of Technology, Kanpur (India); Saxena, Anupam [Department of Mechanical Engineering, Indian Institute of Technology, Kanpur (India); Kumar, Manoj; Kumar, Ajai [Institute for Plasma Research (India)

    2014-11-15

    Graphical abstract: - Highlights: • Improved algorithm works equally well for central as well as for peripherical plasma regions. • Entropy optimized smoothening parameters eliminate user dependencies. • Real time fusion grade plasma diagnostics images. - Abstract: This visible light computerized tomography exercise is a part of a project to establish an auxiliary imaging method to assist other imaging facilities at the Institute of Plasma Research (IPR), India. Space constraints around Aditya Tokamak allow only two orthogonal ports. Each port has one detector array (64 sensors) sensitive to the visual spectrum emitted by H{sub α} emission. The objective here is to report the developments on limited view tomography for hot plasma imaging. Spatially filtered entropy maximization algorithm with non-uniform discretization grids is employed. Estimation of unique kernel smoothening parameters (mask size and exponent factor) depends on entropy function and projection data. It removes requirement of any arbitrary/user-based decision for choosing a regularization factor thus minimizes the chance for biasedness or errors. Synthetic projection data is used to analyse the performance of this modification. The error band in the process of recovery remains under acceptable level (less than 15%) irrespective of the origin of the emissions from the core. Reconstructed hot plasma images/profiles from Aditya Tokamak are shown. These profiles may improve the current understanding about (a) plasma–wall interaction or edge plasma turbulence, (b) control and generation of plasma and (c) correlations between theoretical and engineering advancements in Tokamak reactors.

  17. Plasma diagnostics at Aditya Tokamak by two views visible light tomography

    Graphical abstract: - Highlights: • Improved algorithm works equally well for central as well as for peripherical plasma regions. • Entropy optimized smoothening parameters eliminate user dependencies. • Real time fusion grade plasma diagnostics images. - Abstract: This visible light computerized tomography exercise is a part of a project to establish an auxiliary imaging method to assist other imaging facilities at the Institute of Plasma Research (IPR), India. Space constraints around Aditya Tokamak allow only two orthogonal ports. Each port has one detector array (64 sensors) sensitive to the visual spectrum emitted by Hα emission. The objective here is to report the developments on limited view tomography for hot plasma imaging. Spatially filtered entropy maximization algorithm with non-uniform discretization grids is employed. Estimation of unique kernel smoothening parameters (mask size and exponent factor) depends on entropy function and projection data. It removes requirement of any arbitrary/user-based decision for choosing a regularization factor thus minimizes the chance for biasedness or errors. Synthetic projection data is used to analyse the performance of this modification. The error band in the process of recovery remains under acceptable level (less than 15%) irrespective of the origin of the emissions from the core. Reconstructed hot plasma images/profiles from Aditya Tokamak are shown. These profiles may improve the current understanding about (a) plasma–wall interaction or edge plasma turbulence, (b) control and generation of plasma and (c) correlations between theoretical and engineering advancements in Tokamak reactors

  18. Real-time horizontal position control for Aditya-upgrade tokamak

    Position of plasma column is required to be controlled in real time for improved operation of any tokamak. A PID based system for real-time horizontal plasma position control has been designed for Aditya Upgrade tokamak. Modelling of transfer functions of actuators, plasma and diagnostic system are carried out for ADITYA-U tokamak. The PID controller is optimized using MATLAB-SIMULINK for horizontal position control. Further feed-forward loop is implemented where disturbance due to density variation is suppressed, which results in improved performance as compared to conventional PID operation. In this paper the detailed design of the whole system for real time control of plasma horizontal position in Aditya Upgrade tokamak is presented. (author)

  19. Equilibrium reconstruction of plasma discharges in the Aditya Tokamak

    External magnetic measurements with flux loops and magnetic pick-up coils in tokamaks have provided vital information on the shape of the plasma column and also global current profile parameters, such as the sum of the poloidal beta (βp) and the internal inductance (ℓi). Such a reconstruction needs to be fast and sufficiently accurate such that it can be used routinely as a complementary input with other experimentally measured parameters for any sort of physics analysis of the plasma discharges. Here we present a method which can be used to proficiently reconstruct the current profile parameters, the plasma shapes, and a current density profile satisfying the MHD equilibrium constraint, reasonably conserving the external magnetic measurements. A Grad-Shafranov (GS) equation solver, named as IPREQ, has been developed in IPR to search for the best-fit current density profile. GS equation is a nonlinear elliptical differential equation describing axisymmetric toroidal equilibria. Ohmic transformer current (OT), vertical field coil current (BV) along with the plasma pressure (p) and current (Ip) profiles are used as inputs to the IPREQ code to reconstruct the equilibrium and the poloidal flux, plasma shape, βp and the safety factor (q) are inferred. At the four corners of the square cross-section vacuum vessel of Aditya, there are four magnetic pick-up coils aligned to measure the poloidal magnetic field (Bθ) during a plasma discharge. Further, there are two toroidal flux loops at the shadow of the limiter on the high field side to measure the loop voltage inside the vacuum vessel. Vacuum shots with OT and BV and no fill gas are used to calibrate these coils and loops. Measurement from these coils and flux loops are used to reconstruct the equilibrium consistently with the peak density and temperature measurements. Finally, the reconstructed equilibria are validated against the visible images from the fast visible imaging diagnostic on Aditya. (author)

  20. Development of infrared imaging video bolometer for the ADITYA tokamak

    The Infrared Imaging Video Bolometer (IRVB) is one of the modern plasma imaging diagnostics which provides the measurement of the temporally as well as spatially resolved (2-D/3-D) power profile radiated from plasma devices. The technique has successfully been tested on a large size tokamak (JT-60U) and the same technique is for the first time being utilized for the medium size tokamak ADITYA (R = 75 cm, a = 25 cm, Ip = 80 kA, Te(0) ∼ 350 eV, ∼ 1.5 × 1013 cm3, BT = 0.7 T), where the plasma shot duration is ∼100 ms and radiated power brightness level is ∼2 W/cm2. The diagnostic is utilizing a 6.4 cm×6.4 cm size and 2.5 µm thick, free standing Platinum foil. A square aperture 0.7 × 0.7 cm2 of pinhole camera geometry can provide 9 × 9 bolometer pixel arrays (81 channels) and ∼7 cm of spatial resolution at plasma mid-plane with a 45deg × 45deg wide field of view. This wide field of view covers two semi-tangential views, on either side of the radial view in the tokamak along with a poloidal view. A medium wave infrared camera having 320×240 focal plane array, 200 Hz frame rate, noise equivalent temperature difference ∼20 mK is used and 10 ms of optimal temporal resolution is experimentally achieved. The present paper discusses the design, development and calibration of the system. The performance of the IRVB system for its time response is experimentally investigated and has also been reported here. (author)

  1. Design of high resolution spectroscopic diagnostics for SST-1 and Aditya-U tokamak

    High Resolution spectroscopic diagnostics are proposed for SST-1 and ADITYA-U Tokamak for the measurement of plasma rotation and ion temperature using line radiations emitted by impurity ions. A high resolution Charge eXchange Recombination Spectroscopy (CXRS) using line emission from C VI (n=8◊7) at 529 nm is proposed for SST-1 Tokamak. SST-1 Tokamak is equipped with a heating neutral beam of 40 keV energy with a beam power of 1.2 MW for the measurement of impurity rotation and temperature. The CXRS diagnostic for SST-1 will have a high spatial resolution of ∼ 1cm and a high time resolution of ∼5ms. Imaging X-ray crystal spectroscopy diagnostic (XCS) is proposed for ADITYA-U Tokamak to provide spatially and temporally resolved measurement of plasma rotation and impurity ion behavior. The spectrometer will consist of a spherically bent crystal and CCD detector to measure Ne IX line emission at 13.4474 Å (w) in the toroidal plane of the vacuum vessel with spatial resolution of ∼ 2.8 cm. The diagnostic will provide multiple line of sight measurement to estimate toroidal rotation velocity profile and understand impurity transport for ADITYA-U plasma. Feasibility study for the design of the CXRS diagnostic including a detailed calculation of the photon budget and Etendue budget is presented in this article. Moreover, details of the XCS diagnostic design and system integration with ADITYA-U tokamak are also presented. (author)

  2. Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

    Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

  3. Second-harmonic ion cyclotron resonance heating scenarios of Aditya tokamak plasma

    Asim Kumar Chattopadhyay; S V Kulkarni; R Srinivasan; Aditya Team

    2015-10-01

    Plasma heating with the fast magnetosonic waves in the ion cyclotron range of frequencies (ICRF) is one of the auxiliary heating schemes of Aditya tokamak. Numerical simulation of second-harmonic resonance heating scenarios in low-temperature, low-density Aditya plasma has been carried out for fast magnetosonic wave absorption in ICRF range, using full-wave ion cyclotron heating code TORIC combined with Fokker–Planck quasilinear solver SSFPQL and the results are explained. In such low-temperature, low-density plasma, ion absorption for second-harmonic resonance heating is less but significant amount of direct electron heating is observed.

  4. Thermal electron cyclotron emission measurement on the Aditya tokamak by radiometers

    Thermal electron cyclotron emission (ECE) is measured on a medium size Aditya tokamak by a multi-channel Ka-band radiometer and another multi-channel E-band radiometer. The optically thick second harmonic Ka-band radiometer measured signal is affected by the right-hand cutoff effect beyond ∼25 ms. Due to this cutoff, the electron temperature cannot be measured beyond this time. The plasma density is evaluated for the cutoff frequency channel. It is not possible to also determine the electron temperature from the third harmonic optically thin E-band measurements. Yet these measurements are useful to study sawtooth oscillation phenomena. The sawtooth period and amplitude dependence on measurable plasma parameters are determined and new scaling laws are established for Aditya plasma sawtooth. The propagation delays of inverse sawtooth at different radial channels are used to determine thermal diffusivity. The measured diffusivity (χeHP ∼ 20-31 m2s-1) is found and compared with χePB, which is determined from power balance of background Aditya plasma. The ratio χeHP/χePB is 2-3 for the Aditya plasma discharge. This ratio is comparable with a previous study of heat diffusion on medium size tokamaks

  5. Measurement of electron temperature profile using absorption foil technique for ADITYA Tokamak discharges

    Soft X-Ray imaging array system installed in Aditya tokamak is useful for study the characteristics of sawtooth oscillation, major disruption, Magneto Hydro Dynamic (MHD) activity, and measurement of electron temperature. In most of the tokamaks electron temperature has been calculated using the absorption foil method developed F.C. Jahoda et al. Soft X-Ray imaging system consists of two array silicon surface barrier detectors (SBD) modified for the measurement of chord average electron temperature profile. In this paper, we are first time reporting, temporal and spatial measurement of chord averaged electron temperature (Te) for five different radial positions. In most of Aditya plasma discharges, radial profile of Te is very close to parabolic in nature. Details of experiment and plasma parameter will be discussed. (author)

  6. Study of impurities in Aditya Tokamak during different conditions using quadrupole mass analyzer

    In fusion devices, e.g., Tokamak, the presence of the impurities, i.e. gas species other than the fuel gas, deteriorates plasma and makes confinement difficult. The gas molecules tend to get adsorbed on the surfaces of the solid state materials of the vessel wall during discharges. A Residual Gas Analyzer (RGA) is the most commonly useful instrument to measure the presence and quantity of the various gases in a vacuum system. Quadrupole Mass Analyzer (QMA) is installed on Aditya Tokamak to measure the concentrations of various gas species present in Aditya vacuum system. It is also used to monitor impurities generated during various phases of discharges in Aditya Tokamak. The impurities are reduced by various types of discharge cleaning and in-situ coatings. Presence of residual gas concentration in vacuum system creates limitation for achievement of ultrahigh vacuum and also affects plasma performance. The presence of residual gases is due to different reasons like atmospheric concentration, contamination of the wall materials, outgassing from the exposed materials, permeation, real and virtual leaks

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

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

  8. Divertor coil power supply in Aditya Tokamak for improved plasma operation

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a Tokamak with divertor configuration. This moderate field Tokamak is capable of producing 250 kA of plasma current with 300 ms duration. Two new sets of diverter coils will be added to the system with an objective of producing double null plasmas in Aditya Upgrade Tokamak. Diverter coils, made up of continuously transposed conductor, are low voltage high current carrying poloidal field coils. One set of inner divertor coil has radius of 460 mm containing 6 turns and the other set of 1075 mm radius coil with 1 turn makes the outer divertor coils. The simulated plasma double null equilibrium demands 150 kAT of NI for the inner divertor coils and 10 - 20 kAT of NI for outer divertor coils. To energize the divertor coils with required power, a pulsed DC power supply of 3 MW (100V, 30 kA) has been designed. The designed pulsed DC power supply will be a 3-phase, 12-pulse rectifier based convertor power supply having a duty cycle of 300 ms on-time and 15 minutes off-time. The current rise time in the divertor coils will be ∼ 0.6 MA/sec. Detailed design of the divertor power supply with active controls for real time control of the plasma shape will be discussed in this paper. (author)

  9. Investigation of oxygen impurity transport using the O4+ visible spectral line in the Aditya tokamak

    Intense visible lines from Be-like oxygen impurity are routinely observed in the Aditya tokamak. The spatial profile of brightness of a Be-like oxygen spectral line (2p3p 3D3–2p3d 3F4) at 650.024 nm is used to investigate oxygen impurity transport in typical discharges of the Aditya tokamak. A 1.0 m multi-track spectrometer (Czerny–Turner) capable of simultaneous measurements from eight lines of sight is used to obtain the radial profile of brightness of O4+ spectral emission. The emissivity profile of O4+ spectral emission is obtained from the spatial profile of brightness using an Abel-like matrix inversion. The oxygen transport coefficients are determined by reproducing the experimentally measured emissivity profiles of O4+, using a one-dimensional empirical impurity transport code, STRAHL. Much higher values of the diffusion coefficient compared with the neo-classical values are observed in both the high magnetic field edge region (Dinboardmax∼30 m2 s-1) and the low magnetic field edge region (Doutboardmax∼45 m2 s-1) of typical Aditya ohmic plasmas, which seems to be due to fluctuation-induced transport. The diffusion coefficient at the limiter radius in the low-field (outboard) region is typically ∼ twice as high as that at the limiter radius in the high-field (inboard) region. (paper)

  10. Estimation of effective responsivity of AXUV bolometer in ADITYA tokamak by spectrally resolved radiation power measurement

    The radiation emission from ADITYA Tokamak is routinely measured using AXUV bolometers and the total radiation power loss is estimated from these measurements assuming constant responsivity. This assumption is valid for the current flattop phase of the discharge, where the contribution from long wavelength radiation (> 620 Å) is expected to be small and the AXUV responsivity is almost constant. It is likely that in disruptive discharges, with significant edge radiation, a part of the unaccounted power is in the long wavelength range. A better approach is to experimentally determine an effective responsivity by spectrally resolving the radiation power loss and assigning appropriate weights to spectral ranges. For this purpose, we have installed a multichannel filtered bolometer camera in ADITYA Tokamak. The wide angle view camera houses three single channel AXUV bolometers, of which two view the plasma through different ultraviolet filters and one has an unfiltered view. All the bolometers have the same poloidal view and are located adjacently in the toroidal direction. The initial results of the spectrally resolved bolometer measurements show that the radiation in the spectral range > 1200 Å is significant fraction of the total radiation during the disruptive phase, but doesn't contribute much during the flattop region. An effective average responsivity has been estimated for AXUV bolometer for ADITYA. (author)

  11. Multidirectional plasma flow measurement by Gundestrup Probe in scrape-off layer of ADITYA tokamak

    Sangwan, Deepak; Jha, Ratneshwar; Tanna, Rakesh L. [Institute for Plasma Research, Gandhinagar, Gujarat 382428 (India)

    2015-11-15

    Multidirectional plasma flow measurements by using Gundestrup Probe in the scrape-off layer of ADITYA tokamak are presented. The ADITYA Gundestrup Probe-head consists of eight plates arranged around the ceramic rod and three pins normal to side plates. Plates are used to measure both parallel and perpendicular flows simultaneously and pins are used to measure plasma density and floating potential. A comparison of direct perpendicular flow measurement and by two other plates of Gundestrup Probe is presented. Possible causes of perpendicular flows are identified and compared with the measured flows. It is observed that the mechanism of the parallel flow and the perpendicular flow is different only at high parallel Mach number. A puff of the working gas is used to study its effect on the perpendicular flows and its reversal with the gas puff is observed.

  12. Development of non-circular metal seal for Aditya Tokamak upgrade vacuum vessel

    Existing Aditya Tokamak is being upgraded into a machine with divertor operation. To accommodate divertor magnet coils, existing vacuum vessel will be replaced with new circular section vacuum vessel having volume of ∼1.5 m3. This vacuum vessel is fabricated by SS 304L and can be baked upto 150 °C. The ultimate vacuum envisaged in the vessel is ∼10-9 torr. The vacuum vessel has 112 ports opening of various sizes and shapes, viz. circular, rectangular and triangular types. The circular ports are vacuum sealed using CF metal seal, while the non-circular ports are sealed using metal wire-seals. Customized shaped aluminium wire seals are designed and fabricated for new vacuum vessel. The designed and fabricated aluminium wire seals are tested on local set up in laboratory to confirm its validation as appropriate metal seal for new vacuum vessel for Aditya Tokamak Upgrade. The challenging task of achieving a leak rate less than ∼10-9 torr-l/s with baking upto 150 °C is successfully accomplished on the test bench. The same wire-seals are then successfully used in Aditya Upgrade vessel achieving a base vacuum ∼ 10-9 torr. The flanges with wire seals are required to be tightened specific torque range (25 - 35 N-m) to obtain optimum symmetrical sealing. The wire seals are fabricated in-house using butt welding machine and the stiffness of joints are checked using radiography. This paper presents design, fabrication technique and test results of the wire-seals successfully used in ultra-high vacuum vessel of Aditya Upgrade. (author)

  13. Parametric study of total radiation power loss from the Aditya tokamak using infrared imaging video bolometer

    Infrared Imaging Video Bolometer (IRVB) is a new type of total radiation power loss measurement technique which provides the time resolved two-dimensional images of the line integrated plasma radiation with wide field of view. An IRVB system has been designed, developed, calibrated and tested for its performance and is to be installed on the ADITYA tokamak. This ADITYA IRVB has a broad radiation absorption band ∼1 eV to 85 keV, wide Field of View 46° x 46°, 9 x 9 bolometer pixel array (81 channels), data acquisition rate 166 Hz with a spatial resolution at plasma mid plane of ∼ 7 cm and the Noise Equivalent Power Density (NEPD) ∼200 μW/cm2. Using the IRVB, 2-D radiation brightness images were obtained and analyzed. The present paper describes IRVB data analysis scheme and estimation of total radiation power loss from the ADITYA plasma. Parametric variations of the total radiated power loss obtained from analyzed IRVB images with density, temperature (Te) and plasma current (Ip) had have been reported here. It is found that during plasma current flat-top the total radiation power loss varies from 20% to 40% of the total input ohmic power for different plasma discharges. Also, the radiated power fraction f∼Prad/Pin has been found to be increasing with the increasing average plasma density and decreases with increasing Te and Ip . The recent results also confirm the previous measurements carried out on the ADITYA tokamak using AXUV-Bolometer. (author)

  14. A fixed frequency reflectometer to measure density fluctuations at Aditya Tokamak

    Amongst modern diagnostics of fusion plasmas, microwave methods, both passive and active, play an important role. Microwave Reflectometer is used to measure the plasma density and its fluctuations in fusion research device like tokamak. A fixed frequency (O - mode) microwave reflectometer at 22 GHz (cut - off density nc = 6 X 1012 cm-3) has been designed, developed and used to measure the critical density layer and its fluctuations in Aditya. It can measure the plasma density fluctuations from r = 11 to 22 cm for central electron density 7.5 X 1012 cm-3 and more, respectively. The output signal of reflectometer is analyzed and compared with the density measurement from the microwave interferometer. When the density measured by interferometer is constant, then the fluctuations of local density are seen from the reflectometer signal. Analysis of initial results show that density fluctuation at r = 21 cm in the main plasma has correlation time of about 8 μsec and frequency spectrum is broad. Use of 22 GHz incident wave allows the observation of density fluctuation with wave number in the range of 0 - 9.2 cm-1 from the reflecting region at the receiving horn. Radial variation of the fluctuation level is observed from 5% to 22% for minor radius 11 to 22 cm, respectively. (author)

  15. FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak

    Suratia, Pooja, E-mail: poojasuratia@yahoo.com [Electrical Engineering Department, Faculty of Technology and Engineering, The Maharaja Sayajirao University of Baroda, Kalabhavan, Vadodara 390001, Gujarat (India); Patel, Jigneshkumar, E-mail: jjp@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Rajpal, Rachana, E-mail: rachana@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Kotia, Sorum, E-mail: smkotia-eed@msubaroda.ac.in [Electrical Engineering Department, Faculty of Technology and Engineering, The Maharaja Sayajirao University of Baroda, Kalabhavan, Vadodara 390001, Gujarat (India); Govindarajan, J., E-mail: govindarajan@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer Evaluation and comparison of the working performance of FLC is done with that of PID Controller. Black-Right-Pointing-Pointer FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. Black-Right-Pointing-Pointer FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. Black-Right-Pointing-Pointer Developed FLC controller is able to maintain the plasma column within required range of {+-}0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional-Integral-Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).

  16. FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak

    Highlights: ► Evaluation and comparison of the working performance of FLC is done with that of PID Controller. ► FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. ► FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. ► Developed FLC controller is able to maintain the plasma column within required range of ±0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional–Integral–Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).

  17. Estimation of spectrally resolved total radiation power loss in Aditya Tokamak and its comparison with experimental measurements

    The radiation power loss in Aditya tokamak is routinely measured using AXUV diodes. Both single channel and arrays of AXUV diode are used for the measurement. In addition, filtered channels are used for the measurement of spectrally resolved radiation loss in the VUV region and to estimate the effective responsivity in the operation regimes where there is a significant contribution of lower energy radiation to the total power loss. In the present work, the steady state radiation power loss in Aditya tokamak is modeled using one dimensional impurity transport code, STRAHL under the assumption of toroidal and poloidal symmetries of the plasma. For this purpose, photon emissivity coefficients from ADAS database of the main impurities, such as carbon and oxygen, have been used to estimate the spectrally resolved radiation power loss. The simulated radiation power loss is compared to the experimentally measured radiation power loss from a typical Aditya plasma discharge and the similarities and discrepancies are discussed. (author)

  18. Development of gas puffing system for LHCD experiment in Aditya tokamak

    Lower hybrid (LH) wave coupling experiments have been successfully carried out in Aditya tokamak using 120 kW, pulsed LHCD system based at 3.7 GHz. To enhance the coupling of LH waves in the edge plasma region, an especially designed gas puffing system is developed to inject Hydrogen gas from the electron side of the grill antenna. The developed new gas puffing system consists of a multi-hole gas injection manifold with precisely fabricated holes. The dimensions of the manifold are determined so as to spread the gas uniformly in front of antenna. We achieved precise control of neutral gas injection near the antenna by this new gas puffing system of LHCD as observed by the images taken by fast camera. The gas puff using the manifold near the LH antenna led to considerable reduction in the reflection co-efficient of LH power indicating enhance absorption in plasma. The number of particles injection through gas puffing system has been estimated to figure out the optimum LH power coupling in Aditya tokamak. This paper presents detail of the developed gas puffing system for LHCD experiments and its implication on LHCD experiments. (author)

  19. A set-up for a biased electrode experiment in ADITYA Tokamak

    An experimental set-up to investigate the effect of a biased electrode introduced in the edge region on ADITYA tokamak discharges is presented. A specially designed double-bellow mechanical assembly is fabricated for controlling the electrode location as well as its exposed length inside the plasma. The cylindrical molybdenum electrode is powered by a capacitor-bank based pulsed power supply (PPS) using a semiconductor controlled rectifier (SCR) as a switch with forced commutation. A Langmuir probe array for radial profile measurements of plasma potential and density is fabricated and installed. Standard results of improvement of global confinement have been obtained using a biased electrode. In addition to that, in this paper we show for the first time that the same biasing system can be used to avoid disruptions through stabilisation of magnetohydrodynamic (MHD) modes. Real time disruption control experiments have also been carried out by triggering the bias-voltage on the electrode automatically when the Mirnov probe signal exceeds a preset threshold value using a uniquely designed electronic comparator circuit. Most of the results related to the improved confinement and disruption mitigation are obtained in case of the electrode tip being kept at ∼3 cm inside the last closed flux surface (LCFS) with an exposed length of ∼20 mm in typical discharges of ADITYA tokamak. (paper)

  20. Timing control circuit for real-time control of events in Aditya Tokamak

    Tokamak plasma is prone to many random events having potential for causing severe damages to the machine, such as disruptions, production and elimination of high-energy runaway electrons etc. These events can be mitigated by obtaining pre-cursor signal leading to these events and then taking proper measures just before their onset to avoid their happenings, like disruptions can be mitigated by massive gas injection or putting a bias voltage on an electrode placed inside the plasma, the runaways can be mitigated by gas injection and by applying specific magnetic fields. Hence for real time control of these events, the pre-cursors should be electronically recorded and the mitigation techniques should be initiated by sending triggers to their individual operational systems. To implement these methodologies of real-time controlling of events in Aditya Tokamak, a low cost multi-channel Micro-Controller based timing circuit is designed and developed in-house. This circuit first compares the precursor signals fed into it with the pre-set values and gives a trigger output whenever the signals overshoot the pre-set values. The circuit readies itself for operation along with start of the tokamak discharge and waits up to an initial pre-determined delay and then initiates a trigger at the time of overshooting of precursor signal. The circuit is fully integrated and assembled in compact enclosure with local LCD for threshold and initial trigger-delay monitoring and indicators for full stand-alone operation. The system has been successfully tested in the disruption control by biasing electrode experiments in Aditya tokamak. The MHD oscillations, precursor in this case, is monitored by this circuit and whenever the amplitude of these oscillations overshoot a particular pre-set value, a trigger is generated and delivered to a SCR switch which triggers the voltage on the electrode placed inside the plasma to avoid disruptions. The detailed design features and results will be

  1. A PMT array based diagnostics to measure spatial and temporal behavior of Hα emission from Aditya Tokamak

    The detailed information on fast changing plasma behavior during the breakdown and start-up phase of a tokamak plasma is very essential for achieving good plasma current flat-top region. A Photo multiplier tube (PMT) array based spectroscopic diagnostics has been designed and developed to measure the spatial profile of Hα, Hβ and C III radiation from Aditya tokamak plasma with very fast time response ∼100 μs and also with a good spatial resolution ∼ 3.5 cm at plasma mid plane. The system has been installed on Aditya tokamak to study the breakdown location by monitoring the Hα emission during the plasma formation stage. Two 8 channels linear multi anode PMT arrays with high gains, wide dynamic range and low noise are used as detector. The module comes with built-in high voltage power supply and built-in amplifier. Collimated light has been collected from the plasma along 16 line-of-sights passing through the entire plasma poloidal cross section from the top port of Aditya tokamak and transferred to the PMT array using 1 mm core diameter optical fibers. The Hα spectra is obtained using 8 miniature interference filters (IF) centered at 656.3 nm placed in front of the PMT array. For the 2nd PMT array, another arrangement for wavelength selection is developed using bigger 2.5” IF, where lights from multiple fibers can be passed through for wavelength selection simultaneously. The spatial and temporal profiles of Hα emissions have been studied during the formation phase of Aditya tokamak plasma by changing the vertical field and delay of its application with respect to loop voltage. It was found that the plasma initiates in the high field side of tokamak most of the times. The details on experimental set-up and the results of the experiments will be discussed in this presentation. (author)

  2. Comparison of different atomic databases used for evaluating transport coefficients in Aditya Tokamak

    Oxygen impurity transport in typical discharges of Aditya tokamak has been estimated using spatial profile of brightness of Be-like oxygen (O4+) spectral line (2p3p 3D3-2p3d 3F4) at 650.024 nm. This O4+ spectrum is recorded using a 1.0 m multi-track spectrometer (Czerny-Turner) capable of simultaneous measurements from eight lines of sights. The emissivity profile of O4+ spectral emission is obtained from the spatial profile of brightness using an Abel-like matrix inversion. The oxygen transport coefficients are then determined by reproducing the experimentally measured emissivity profiles of O4+, using a one-dimensional empirical impurity transport code, STRAHL. To calculate the emissivity, photon emissivity coefficient (PEC) is required along with electron and O4+ density, which is the output of STRAHL. The PEC values depend on both electron density and temperature and are obtained from ADAS and NIFS atomic databases. Using both the databases, much higher values of diffusion coefficient compared to the neo-classical values are observed in the high and low magnetic field edge regions of typical Aditya Ohmic plasmas. The obtained values of diffusion coefficients using PEC values from both the databases are compared with the diffusion coefficients calculated from the fluctuation induced transport in both the inboard and outboard edge regions. Although similar profiles for diffusion coefficients are obtained using PEC values from both databases, the magnitude differs considerably. (author)

  3. Measurement of spatial and temporal behavior of Hα emission from Aditya tokamak using a diagnostic based on a photomultiplier tube array

    Chowdhuri, M. B.; Ghosh, J.; Manchanda, R.; Kumar, Ajay; Banerjee, S.; Ramaiya, N.; Virani, Niral; Mali, Aniruddh; Amardas, A.; Kumar, Pintu; Tanna, R. L.; Gupta, C. N.; Bhatt, S. B.; Chattopadhyay, P. K.

    2014-11-01

    A photo multiplier tube (PMT) array based spectroscopic diagnostic with fast time response of 10 μs and spatial resolution ˜3 cm has been developed and installed on Aditya tokamak to study the spatial and temporal behavior of Hα emissions from typical discharges. Collimated light has been collected from the plasma along 16 lines of sight passing through entire plasma poloidal cross section of Aditya and detected by two 8 channels PMT arrays after selecting Hα emission using interference filter. The studies are carried out during plasma formation phase of Aditya by changing vertical field and its delay with respect to loop voltage. It is observed that plasma initiated in the high field side in typical discharges of Aditya. The plasma formation position is matched with null field location estimated through simulation.

  4. Measurement of spatial and temporal behavior of H(α) emission from Aditya tokamak using a diagnostic based on a photomultiplier tube array.

    Chowdhuri, M B; Ghosh, J; Manchanda, R; Kumar, Ajay; Banerjee, S; Ramaiya, N; Virani, Niral; Mali, Aniruddh; Amardas, A; Kumar, Pintu; Tanna, R L; Gupta, C N; Bhatt, S B; Chattopadhyay, P K

    2014-11-01

    A photo multiplier tube (PMT) array based spectroscopic diagnostic with fast time response of 10 μs and spatial resolution ∼3 cm has been developed and installed on Aditya tokamak to study the spatial and temporal behavior of Hα emissions from typical discharges. Collimated light has been collected from the plasma along 16 lines of sight passing through entire plasma poloidal cross section of Aditya and detected by two 8 channels PMT arrays after selecting Hα emission using interference filter. The studies are carried out during plasma formation phase of Aditya by changing vertical field and its delay with respect to loop voltage. It is observed that plasma initiated in the high field side in typical discharges of Aditya. The plasma formation position is matched with null field location estimated through simulation. PMID:25430318

  5. Measurement of spatial and temporal behavior of Hα emission from Aditya tokamak using a diagnostic based on a photomultiplier tube array

    A photo multiplier tube (PMT) array based spectroscopic diagnostic with fast time response of 10 μs and spatial resolution ∼3 cm has been developed and installed on Aditya tokamak to study the spatial and temporal behavior of Hα emissions from typical discharges. Collimated light has been collected from the plasma along 16 lines of sight passing through entire plasma poloidal cross section of Aditya and detected by two 8 channels PMT arrays after selecting Hα emission using interference filter. The studies are carried out during plasma formation phase of Aditya by changing vertical field and its delay with respect to loop voltage. It is observed that plasma initiated in the high field side in typical discharges of Aditya. The plasma formation position is matched with null field location estimated through simulation

  6. Measurement of spatial and temporal behavior of H{sub α} emission from Aditya tokamak using a diagnostic based on a photomultiplier tube array

    Chowdhuri, M. B., E-mail: malay@ipr.res.in; Ghosh, J.; Manchanda, R.; Banerjee, S.; Ramaiya, N.; Virani, Niral; Mali, Aniruddh; Amardas, A.; Kumar, Pintu; Tanna, R. L.; Gupta, C. N.; Bhatt, S. B.; Chattopadhyay, P. K. [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat 382 428 (India); Kumar, Ajay [Metallurgical Engineering and Material Science Department, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India)

    2014-11-15

    A photo multiplier tube (PMT) array based spectroscopic diagnostic with fast time response of 10 μs and spatial resolution ∼3 cm has been developed and installed on Aditya tokamak to study the spatial and temporal behavior of H{sub α} emissions from typical discharges. Collimated light has been collected from the plasma along 16 lines of sight passing through entire plasma poloidal cross section of Aditya and detected by two 8 channels PMT arrays after selecting H{sub α} emission using interference filter. The studies are carried out during plasma formation phase of Aditya by changing vertical field and its delay with respect to loop voltage. It is observed that plasma initiated in the high field side in typical discharges of Aditya. The plasma formation position is matched with null field location estimated through simulation.

  7. Study of pellet fuelling requirements for Aditya and SST-1 Tokamak

    In last few decades, pellet injection has become an important tool for fuelling high temperature plasma. In this regard, pellet injection related scenarios in SST-1 and Aditya tokamak plasma is presented in this paper. Considering the density limit of plasma in various operational parameters, cylindrical pellet (equal aspect ratio) of different sizes is chosen for this purpose. With regard to the ablation rate of the pellet in plasma, for core penetration, the injection speed is decided to be <500 m/s. Single pellet injector system (SPINS) developed at IPR will be installed for this purpose. The proposed injector is an in-situ light gas gun type injector, where a pellet is accelerated to higher speed by using high pressure propellant gas. At present, a cylindrical pellet size of 4 to 5 mm and speed ranging from 600 - 900 m/s has been achieved in test bench operation. In the early phase, pellet induced plasma disruption studies by injecting pellets from a radial outboard location have been planned. (author)

  8. Instability analysis in Aditya tokamak discharges with the help of soft x-ray

    Sawtooth oscillations (internal disruptions) and major disruptions are routinely observed in ohmically heated Aditya tokamak discharges. Soft x-ray (SXR) tomography has been used as the main tool to analyse the instabilities in the tokamak discharges along with other supportive diagnostics. SXR tomography is done with the help of a single array of detectors assuming rigid rotation of the modes to analyse the mode structure of internal disruption. The dominant frequencies obtained by the fast Fourier transform (FFT) analysis of the signal at the time of internal disruption are the harmonics of the same mode which are common in toroidal system. The presence of such harmonics makes the signal non-sinusoidal and could easily couple in resonance with the mode oscillations at higher q-surfaces to accelerate the major disruption. The growing m/n=1/1 oscillation at the time of internal disruption and the tomographic images indicate that the sawtooth instabilities seem to be due to the total reconnection model by Kadomtsev, but the crash time according to Kadomtsev model does not obey the observed experimental value. The m/n=1/1 mode rotation is also clear at the time of internal disruption from the tomographic images. After analysis of all other probable possibilities coupling of m/n=2/1 and m/n=1/1 modes appears to be the main mechanism for the major disruption. Singular value decomposition (SVD) method has been used to analyse the time series of tomographic reconstructions to identify the dominant magnetohydrodynamic modes and to show different features of the spatio temporal evolution of the emissivity distribution. (author)

  9. Silicon drift detector based X-ray spectroscopy diagnostic system for the study of non-thermal electrons at Aditya tokamak.

    Purohit, S; Joisa, Y S; Raval, J V; Ghosh, J; Tanna, R; Shukla, B K; Bhatt, S B

    2014-11-01

    Silicon drift detector based X-ray spectrometer diagnostic was developed to study the non-thermal electron for Aditya tokamak plasma. The diagnostic was mounted on a radial mid plane port at the Aditya. The objective of diagnostic includes the estimation of the non-thermal electron temperature for the ohmically heated plasma. Bi-Maxwellian plasma model was adopted for the temperature estimation. Along with that the study of high Z impurity line radiation from the ECR pre-ionization experiments was also aimed. The performance and first experimental results from the new X-ray spectrometer system are presented. PMID:25430326

  10. Conceptual design of automation of ICRH vacuum system on Aditya Tokamak

    Rathi, Dharmendra; Mishra, Kishore; Joshi, Ramesh; Jadav, H. M.; Kulkarni, S. V.; ICRH-RF Group

    2010-02-01

    Ion Cyclotron Resonance Heating (ICRH) is a successful heating method for a fusion device due to a localized power deposition profile and is an established technology for raising temperature of ion species. The ADITYA-ICRH system consists of RF generator, pressurisable transmission line, a matching network, vacuum transmission line section (VTL) and antenna. The intermediate VTL provides vacuum isolation from that of ADITYA at one end and also separates the pressurisable transmission line at the other end. The ICRH vacuum system consists of VTL, a turbo molecular pump (TMP), pneumatic gate valve, ionization gauge, one vacuum window at antenna side and a gas barrier towards other side. During ADITYA-ICRH operation the ICRH vacuum system should be operated remotely with all the necessary safety interlocks, controls and status at the RF control room. In this paper the schematics of automated vacuum system with interlocks, sequence of automation with flow chart and related results will be discussed.

  11. Conceptual design of automation of ICRH vacuum system on Aditya Tokamak

    Ion Cyclotron Resonance Heating (ICRH) is a successful heating method for a fusion device due to a localized power deposition profile and is an established technology for raising temperature of ion species. The ADITYA-ICRH system consists of RF generator, pressurisable transmission line, a matching network, vacuum transmission line section (VTL) and antenna. The intermediate VTL provides vacuum isolation from that of ADITYA at one end and also separates the pressurisable transmission line at the other end. The ICRH vacuum system consists of VTL, a turbo molecular pump (TMP), pneumatic gate valve, ionization gauge, one vacuum window at antenna side and a gas barrier towards other side. During ADITYA-ICRH operation the ICRH vacuum system should be operated remotely with all the necessary safety interlocks, controls and status at the RF control room. In this paper the schematics of automated vacuum system with interlocks, sequence of automation with flow chart and related results will be discussed.

  12. A synthetic diagnostic to modelled expected 2-D radiation power loss profile for the infrared imaging video bolometer of the Aditya tokamak

    A 'synthetic diagnostic' has been developed to theoretically estimate the radiation from the ADITYA tokamak plasma using Infrared Imaging Video Bolometer (IRVB). These theoretical results will then be compared with the results obtained experimentally. The IRVB is a two dimensional (2-D) plasma radiation imaging diagnostic IRVB is used to measure time resolved 2-D profile of radiation power loss with wide field of view (FOV). The synthetic IRVB assumes symmetry in the tokamak. In poloidal cross-section it assumes symmetric parabolic profiles of plasma temperature, plasma density and impurity density. The IRVB system is essentially a pinhole camera system. It traces the line of sights of each bolometer pixel through the plasma volume and calculates local power emissivity at each volume element in space using the radiative cooling rates of plasma impurity. Finally line integrated emissivity 2-D profile provides a brightness profile at each bolometer pixel. This brightness profile is the expected IRVB image at foil location By considering the system etendue the power loss profile can be computed. Using the synthetic diagnostic, spatial response of the experimental diagnostic, FOV, expected signal level and Signal to Noise ratio can be determined. The synthetic IRVB used to simulate ADITYA-IRVB diagnostic and results were compared with experimental results. (author)

  13. Research using small tokamaks

    These proceedings of the IAEA-sponsored meeting held in Nice, France 10-11 October, 1988, contain the manuscripts of the 21 reports dealing with research using small tokamaks. The purpose of this meeting was to highlight some of the achievements of small tokamaks and alternative magnetic confinement concepts and assess the suitability of starting new programs, particularly in developing countries. Papers presented were either review papers, or were detailed descriptions of particular experiments or concepts. Refs, figs and tabs

  14. Research using small tokamaks

    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

  15. Experiments on Tokamak ADITYA

    It is well known that the Greenwald limit is in reality a limit on edge particle confinement that leads to the loss of edge thermal equilibrium. While the radiative collapse is relatively well understood, questions remain about the exact dynamics of convectively driven collapse. We have examined the role of the Molecular Beam Injection (MBI) and the Gas Puff fuelling methods in the determination of the density limit when such a collapse is imminent. It is seen that, broad pulses of MBI, when fired in quick succession, generate a limit close to that in the case of gas-puff. Short pulses with larger separation in time lead to a significantly higher limiting density. Very large turbulent flux (r>) appears just before the collapse along with rapid changes in the scrape-off-layer scalelength for the former cases, unlike the case with smaller, widely spaced MBI pulses. (author)

  16. Joint research using small tokamaks

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

  17. Joint research using small tokamaks

    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)

  18. Ion cyclotron resonance heating system on Aditya

    D Bora; Sunil Kumar; Raj Singh; S V Kulkarni; A Mukherjee; J P Singh; Raguraj Singh; S Dani; A Patel; Sai Kumar; V George; Y S S Srinivas; P Khilar; M Kushwah; P Shah; H M Jadav; Rajnish Kumar; S Gangopadhyay; H Machhar; B Kadia; K Parmar; A Bhardwaj; Suresh Adav; D Rathi; D S Bhattacharya

    2005-02-01

    An ion cyclotron resonance heating (ICRH) system has been designed, fabricated indigenously and commissioned on Tokamak Aditya. The system has been commissioned to operate between 20·0 and 47·0 MHz at a maximum power of 200 kW continuous wave (CW). Duration of 500 ms is sufficient for operation on Aditya, however, the same system feeds the final stage of the 1·5 MW ICRH system being prepared for the steady-state superconducting tokamak (SST-1) for a duration of 1000 s. Radio frequency (RF) power (225 kW) has been generated and successfully tested on a dummy load for 100s at 30·0 MHz. Lower powers have been coupled to Aditya in a breakdown experiment. We describe the system in detail in this work.

  19. Research using small tokamaks

    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

  20. [High beta tokamak research

    Our activities on High Beta Tokamak Research during the past 20 months of the present grant period can be divided into six areas: reconstruction and modeling of high beta equilibria in HBT; measurement and analysis of MHD instabilities observed in HBT; measurements of impurity transport; diagnostic development on HBT; numerical parameterization of the second stability regime; and conceptual design and assembly of HBT-EP. Each of these is described in some detail in the sections of this progress report

  1. Status of tokamak research

    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

  2. Research using small tokamaks

    The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway

  3. Joint research using small tokamaks

    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

  4. Magnetohydrodynamic instability, feedback stabilization, and disruption study for the Korea superconducting tokamak advanced research tokamak

    Passive and active feedback stabilization schemes being considered in Korea Superconducting Tokamak Advanced Research (KSTAR) device for the stabilization of the resistive magnetohydrodynamic modes such as the resistive wall and the neoclassical tearing are briefly introduced. A short summary is also presented on the tokamak simulation results of disruption dynamics and load in the KSTAR tokamak obtained using the tokamak simulation code (TSC)

  5. Tokamak research in the Soviet Union

    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.)

  6. Recent Activities on the Experimental Research Programme Using 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 (CRP) is discussed in this paper. Besides the presentation of the recent activities on the experimental research programme using small tokamaks and scientific results achieved at the participating laboratories, information is provided about the organisation of the co-ordinated research project. Future plans of the co-ordinated activities within the CRP are discussed

  7. Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak

    Maurya, Gulab Singh; Kumar, Rohit; Rai, Awadhesh Kumar, E-mail: awadheshkrai@rediffmail.com [Laser Spectroscopy Research Laboratory, Department of Physics, University of Allahabad, UP 211002 (India); Kumar, Ajai [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat 382428 (India)

    2015-12-15

    In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as “back collection method” to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.

  8. Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak

    In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as “back collection method” to record LIBS spectra of impurities deposited on the inner surface of optical window is presented

  9. Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak.

    Maurya, Gulab Singh; Kumar, Rohit; Kumar, Ajai; Rai, Awadhesh Kumar

    2015-12-01

    In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as "back collection method" to record LIBS spectra of impurities deposited on the inner surface of optical window is presented. PMID:26724011

  10. Operation of ADITYA Thomson scattering system: measurement of temperature and density

    ADITYA Thomson scattering (TS) system is a single point measurement system operated using a 10 J ruby laser and a 1 meter grating spectrometer. Multi-slit optical fibers are arranged at the image plane of the spectrometer so that each fiber slit collects 2 nm band of scattered spectrum. Each slit of the fiber bundle is coupled to high gain Photomultiplier tubes (PMT). Standard white light source is used to calibrate the optical fiber transmission and the laser light itself is used to calibrate the relative gain of the PMT. Rayleigh scattering has been performed for the absolute calibration of the TS system. The temperature of ADITYA plasma has been calculated using the conventional method of estimation (calculated using the slope of logarithmic intensity vs the square of delta lambda). It has been observed that the core temperature of ADITYA Tokamak plasma is in the range of 300 to 600 eV for different plasma shots and the density 2-3 X 1013/cc. The time evolution of the plasma discharge has been studied by firing the laser at different times of the discharge assuming the shots are identical. In some of the discharges, the velocity distribution appears to be non Maxwellian. (author)

  11. Research using small tokamaks. Proceedings of a technical committee meeting

    The technical reports in these proceedings were presented at the IAEA Technical Committee Meeting on research Using Small Tokamaks, held in Ahmedabad, India, 6-7 December 1995. The purpose of this annual meeting is to provide a forum for the exchange of information on various small and medium sized plasma experiments, not only for tokamaks. The potential benefits of these research programmes are to: test theories, such as effects of the plasma rotation; check empirical scalings, such as density limits; develop fusion technology hardware; develop plasma diagnostics; such as tomography; and to train scientists, engineers, technicians, and students, particularly in developing IAEA Member States

  12. Estimation of post disruption plasma temperature for fast current quench Aditya plasma shots

    Characteristics of tokamak current quenches are an important issue for the determination of electromagnetic forces that act on the in-vessel components and vacuum vessel during major disruptions. It is observed that thermal quench is followed by a sharp current decay. Fast current quench disruptive plasma shots were investigated for ADITYA tokamak. The current decay time was determined for the selected shots, which were in the range of 0.8 msec to 2.5 msec. This current decay information was then applied to L/R model, frequently employed for the estimation of the current decay time in tokamak plasmas, considering plasma inductance and plasma resistivity. This methodology was adopted for the estimation of the post disruption plasma temperature using the experimentally observed current decay time for the fast current quench disruptive ADITYA plasma shots. The study reveals that for the identified shots there is a constant increase in the current decay time with the post disruption plasma temperature. The investigations also explore the behavior post disruption plasma temperature and the current decay time as a function of the edge safety factor, Q. Post disruption plasma temperature and the current decay time exhibits a decrease with the increase in the value Q. (author)

  13. Electronic database code upgradation for ADITYA experiments

    Electronic database code processes ADITYA experimental captured raw data to record measured plasma parameters for analysis. Rather than physical channel, flexibly the revised code use unique logical channel number assigned signal raw data and variables to process and produce ensured error-free summarized result comprises calculated value of edge safety factor. (author)

  14. DIII-D Advanced Tokamak Research Overview

    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 βNH 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

  15. Spherical tokamak research for fusion reactor

    Between ITER and the commercial fusion reactor, there are many technological problems to be solved such as cost, neutron and steady-state operation. In the conceptual design of VECTOR and Slim CS reactors it was shown that the key is 'low aspect ratio'. The spherical tokamak (ST) has been expected as the base for fusion reactors. In US, ST is considered as a non-superconducting reactor for use in the neutron irradiation facility. Conceptual design of the superconducting ST reactor is conducted in Japan and Korea independently. In the present article, the prospect of the ST reactor design is discussed. (author)

  16. TIBER: Tokamak Ignition/Burn Experimental Research. Final design report

    The Tokamak Ignition/Burn Experimental Research (TIBER) device is the smallest superconductivity tokamak designed to date. In the design plasma shaping is used to achieve a high plasma beta. Neutron shielding is minimized to achieve the desired small device size, but the superconducting magnets must be shielded sufficiently to reduce the neutron heat load and the gamma-ray dose to various components of the device. Specifications of the plasma-shaping coil, the shielding, coaling, requirements, and heating modes are given. 61 refs., 92 figs., 30 tabs

  17. Abstracts of the International seminar 'Experimental possibilities of KTM tokamak and research programme'

    The International seminar 'Experimental possibilities of KTM tokamak and research programme' was held in 10-12 October 2005 in Astana city (Kazakhstan). The seminar was dedicated to problems of KTM tokamak commissioning. The Collection of abstracts comprises 45 papers

  18. A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

    Mitsuru Kikuchi

    2010-01-01

    Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR) best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fund...

  19. Future directions in fusion research: Super high field tokamaks

    Recent experimental results and advances in magnet engineering suggest that super high field, high aspect ratio tokamak devices could be a very efficient way to achieve burning plasma conditions and could open up a new area of research. Copper magnet devices with fields of 13 to 25 T at the plasma are considered. The super high field approach could also provide advantages for ETR and demonstration/commercial reactor concepts (magnetic fields at the plasma in the 8 to 13 T range)

  20. Overview of Physics Research on the TCV Tokamak

    Alberti, S.; Amorim, P.; Angioni, C.; Andrébe, Y.; Asp, E.; Behn, R.; Bencze, A.; Berrino, J.; Blanchard, P.; Bortolon, A.; Brunner, R.; Camenen, Y.; Coda, S.; Curchod, L.; DeMeijere, K.; Droz, E.; Duval, B.P.; Fable, E.; Fasel, D.; Fasoli, A.; Felici, F.; Furno, I.; Garcia, E.O.; Giruzzi, G.; Gnesin, S.; Goodman, T.; Graves, J.; Gudozhnik, A.; Gulejová, B.; Henderson, M.; Hogge, J. Ph.; Horáček, Jan; Isoz, P. F.; Joye, B.; Karpushov, A.; Kim, S.-H.; Lister, J. B.; Llobet, X.; Madeira, T.; Marinoni, A.; Marki, J.; Martin, Y.; Maslov, M.; Medvedev, S.S.; Moret, J. M.; Paley, J.; Pavlov, I.; Perez, A.; Piffl, Vojtěch; Piras, F.; Pitts, R.A.; Pitzschke, A.; Pochelon, A.; Porte, L.; Reimerdes, H.; Rossel, J.; Sauter, O.; Scarabosio, A.; Schlatter, C.; Sushkov, A.; Testa, D.; Tonetti, G.; Tskhakaya, D.; Tran, M. Q.; Turco, F.; Turri, G.; Tye, R.; Udintsev, V.; Véres, G.; Weisen, H.; Zhuchkova, A.; Zucca, C.

    Geneve: International Atomic Energy Agency, 2008, 1-1-0/OV. [IAEA Fusion Energy Conference/22nd./. Geneva (CH), 13.10.2008-18.10.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak TCV * parallel flow Subject RIV: BL - Plasma and Gas Discharge Physics http://www-pub.iaea.org/MTCD/Meetings/fec2008pp.asphttp://www.fec2008.ch/preprints/ov_1-1.pdf

  1. Overview of physics research on the TCV tokamak

    Fasoli, A.; Alberti, S.; Amorim, P.; Angioni, C.; Asp, E.; Behn, R.; Bencze, A.; Berrino, J.; Blanchard, P.; Bortolon, A.; Brunner, S.; Camenen, Y.; Cirant, S.; Coda, S.; Curchod, L.; DeMeijere, K.; Duval, B. P.; Fable, E.; Fasel, D.; Felici, F.; Furno, I.; Garcia, O.E.; Giruzzi, G.; Gnesin, S.; Goodman, T.; Graves, J.; Gudozhnik, A.; Gulejova, B.; Henderson, M.; Hogge, J. Ph.; Horáček, Jan; Joye, B.; Karpushov, A.; Kim, S.-H.; Laqua, H.; Lister, J. B.; Llobet, X.; Madeira, T.; Marinoni, A.; Marki, J.; Martin, Y.; Maslov, M.; Medvedev, S.; Moret, J.-M.; Paley, J.; Pavlov, I.; Piffl, Vojtěch; Piras, F.; Pitts, R.A.; Pitzschke, A.; Pochelon, A.; Porte, L.; Reimerdes, H.; Rossel, J.; Sauter, O.; Scarabosio, A.; Schlatter, C.; Sushkov, A.; Testa, D.; Tonetti, G.; Tskhakaya, D.; Tran, M. Q.; Turco, F.; Turri, G.; Tye, R.; Udintsev, V.; Véres, G.; Villard, L.; Weisen, H.; Zhuchkova, A.; Zucca, C.

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104005-104005. ISSN 0029-5515 Institutional research plan: CEZ:AV0Z20430508 Keywords : overview highlights * fusion research * tokamak TCV * self-generated current * H-mode physics * Electron internal transport barrier * electron cyclotron heating * electron cyclotron current drive physics * density peaking * MHDactivity * edge physics * reciprocating Mach probe * Pfirsch–Schlueter component. Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://stacks.iop.org/NF/49/104005

  2. Romanian research in the field of Tokamak fusion reactors

    To re-create the conditions of the sun and stars for the production of fusion energy on earth, scientists most accomplish three major tasks. They have already passed the first task by achieving the necessary temperatures. In same cases, they have attained temperatures as high as 510 million degrees, 20 times more then the temperature at the center of the sun. Secondly, they need to demonstrate sustained reactions where substantial amounts of energy are produced. The third major milestone for fusion would be operation of a demonstration fusion power plant. Many different magnetic confined schemes have been studied. The one which is receiving the greatest attention in the international magnetic fusion energy programme is the tokamak concept, and represents actually the most advanced fusion devices. The advantage of fusion are: - abundant fuel supply; - no risk of a nuclear accident; - no air pollution; - no high-level nuclear waste; - no generation of weapons material. The present objectives and research priorities of the fusion community are: - continuation of ongoing research; - concept improvements; - long term technology. Our research programme in the field of tokamak fusion reactions is performed mainly in the frame of international cooperation with 'I.V. Kurchatov' Nuclear Fusion Institute from Moscow, Institute of Applied Mathematics from Grenoble, Research Center from Cadarache, 'Max-Planck' Institute for Plasma Physics from Garching at Munich and Columbia University from New York. The activities carried out under our programme are closely coordinated with those of the European Atomic Energy Community and are related to current problems concerning equilibrium, stability, transport and diagnostics of tokamak plasmas. Our results are mentioned in the International Atomic Energy Agency's World Survey of Activities in Controlled Fusion Research in 1997 and the European Community's Reports EUR FUR BRU from 1993 and 1996. (author)

  3. A brief overview of Tokamak fusion research

    Fusion, the nuclear engine that powers the sun and stars, has been pursued by scientists for decades as the ultimate source of energy. It promises an almost inexhaustible fuel supply with the oceans containing sufficient fusion fuel to outlast the expected life of the sun. Fusion is a process whose waste is inert and whose components know no geographical bounds. Scientists have pondered the laws governing the fusion process since the 1940's, and since the late 1950's laboratory devices have been constructed to test and further develop the theories. To achieve fusion, the joining of light atomic nuclei (as opposed to the splitting of heavy elements in the fission process), the natural tendency of the nuclei to repel each other due to their like electrical charges must be overcome. As the fusion takes place, some of the matter of the nuclei is converted to energy. In the stars fusion is accomplished largely by enormous gravitational forces. On earth the fusion fuel must be heated by other means to increase the energy of the particles to force them to fuse. Therein lies the challenge of fusion research - how to heat sufficient matter to hundreds of millions of degrees and contain it long enough for a controlled and sustained fusion reaction to take place. The method that presently shows the most promise is to contain a plasma (an ionized gas - the fourth state of matter) in a magnetic field while heating the plasma by means of high energy neutral particle beams or radio frequency waves

  4. A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

    Mitsuru Kikuchi

    2010-11-01

    Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

  5. Proposals for an influential role of small tokamaks in mainstream fusion physics and technology research

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confinement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies, which because of the risky nature cannot be done in large machines without preliminary studies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. All these tasks are well recognised and reflected in documents and understood by the large tokamak teams. Recent experimental results will be presented of contributions to mainstream fusion physics and technology research on small tokamaks involved in the IAEA Coordinated Research Project 'Joint Research using small tokamaks', started in 2004

  6. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new

  7. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    Kim, S. K.; Lee, K. W.; Hwang, C. K.; Hong, B. G.; Hong, G. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-05-01

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new.

  8. High-beta tokamak research. Annual progress report, 1 August 1982-1 August 1983

    The main research objectives during the past year fell into four areas: (1) detailed observations over a range of high-beta tokamak equilibria; (2) fabrication of an improved and more flexible high-beta tokamak based on our understanding of the present Torus II; (3) extension of the pulse length to 100 usec with power crowbar operation of the equilibrium field coil sets; and (4) comparison of our equilibrium and stability observations with computational models of MHD equilibrium and stability

  9. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method

  10. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    Kim, Dong-Hwan [Department of Nanoscale Semiconductor Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Hong, Suk-Ho [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); National Fusion Research Institute (NFRI), Daejeon 305-333 (Korea, Republic of); Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook, E-mail: joykang@hanyang.ac.kr [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of)

    2015-12-15

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.

  11. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method.

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-12-01

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method. PMID:26724028

  12. Plasma diagnostics at Aditya Tokomak by two views visible light tomography

    Plasma imaging has always been a requirement for development of correlations between theoretical and engineering advancements in tokomak reactors. Technological constraints do not allow putting sufficient imaging instruments. This visible tomography exercise is a part of a project for establishing an auxiliary imaging method that would assist other surrounding imaging facilities at Institute of Plasma Research (IPR) India. Space constraints around Aditya Tokomak allow only two orthogonal ports. Data measurement is performed using two arrays of 64 detectors that are sensitive to optical spectrum. The two view arrangement is a worst case scenario (as far as number of projections is concerned) but it is not implausible. An algorithm is developed for such limited detector and limited-view tomography cases. Spatial filtered entropy maximization technique is hybridized with adaptive discretization grids to find the best possible solution. Reconstruction using synthetic projection data, similar to the real measurement geometry, shows that significant reduction in r.m.s. error is obtained. Real time plasma images/profiles are reconstructed using multiple shots of thin hot plasma from Aditya Tokomak. These profiles help to understand the real time plasma-wall interaction at different stages of plasma generation due to edge plasma turbulence. It also helps to control the generation of plasma. (author)

  13. Overview of recent and current research on the TCV tokamak

    S. Codathe TCV Team

    2013-10-01

    Through a diverse research programme, the Tokamak à Configuration Variable (TCV) addresses physics issues and develops tools for ITER and for the longer term goals of nuclear fusion, relying especially on its extreme plasma shaping and electron cyclotron resonance heating (ECRH) launching flexibility and preparing for an ECRH and NBI power upgrade. Localized edge heating was unexpectedly found to decrease the period and relative energy loss of edge localized modes (ELMs). Successful ELM pacing has been demonstrated by following individual ELM detection with an ECRH power cut before turning the power back up to trigger the next ELM, the duration of the cut determining the ELM period. Negative triangularity was also seen to reduce the ELM energy release. H-mode studies have focused on the L-H threshold dependence on the main ion species and on the divertor leg length. Both L- and H-modes have been explored in the snowflake configuration with emphasis on edge measurements, revealing that the heat flux to the strike points on the secondary separatrix increases as the X-points approach each other, well before they coalesce. In L-mode, a systematic scan of the auxiliary power deposition profile, with no effect on confinement, has ruled it out as the cause of confinement degradation. An ECRH power absorption observer based on transmitted stray radiation was validated for eventual polarization control. A new profile control methodology was introduced, relying on real-time modelling to supplement diagnostic information; the RAPTOR current transport code in particular has been employed for joint control of the internal inductance and central temperature. An internal inductance controller using the ohmic transformer has also been demonstrated. Fundamental investigations of neoclassical tearing mode (NTM) seed island formation by sawtooth crashes and of NTM destabilization in the absence of a sawtooth trigger were carried out. Both stabilizing and destabilizing agents

  14. 20 years of research on the Alcator C-Mod tokamak

    Greenwald, Martin; Bader, A; Baek, S.; M. Bakhtiari; Barnard, H.; Beck, W.; Bergerson, W; Bespamyatnov, I; Bonoli, P.; Brower, D; Brunner, D.; Burke, W.; Candy, J.; Churchill, M; Cziegler, I.

    2014-01-01

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of crit...

  15. High-beta tokamak research. Annual progress report, August 1, 1983-July 30, 1984

    Our main research objectives during the past year fell into four areas: (1) construction and initial operation of the new tokamak, HBT; (2) further numerical modeling of the Torus II experimental equilibria using the PPPL equilibrium and stability codes; (3) diagnostic development; and (4) ICRF antenna coupling calculation in 2D and rf current drive

  16. Aditya: India’s First Observatory in Space to Study the Sun

    Nandi, Dibyendu

    2015-08-01

    Recognizing the need and advantages of continuous solar observations from space, and to further its goal of supporting scientific and technological advances, the Indian Space Research Organization is planning India’s first space mission to observe the Sun. Nicknamed Aditya, this ambitious project aims to place a comprehensive solar observatory at the Lagrange point L1 which will allow uninterrupted views of the Sun. The diverse set of instruments being planned to fly onboard this mission includes a visible emission line coronagraph, a solar ultraviolet imaging telescope, high- and low-energy X-ray spectrometers, a plasma analyzer and a particle detector package for in-situ measurements. In this talk I will provide a brief overview of these instruments and discuss the science objectives of this mission.

  17. The Recent Research Work on the J-TEXT Tokamak

    Full text: The main results from the J-TEXT tokamak in the last two years, which emphasized the observation and analysis of MHD activity, are summarized and presented in this meeting. Static resonant magnetic perturbations generated by saddle coil currents are applied to J-TEXT Ohmic plasmas in order to study their influence on MHD instabilities. With sufficiently large RMPs, the m/n = 2/1 (m and n are the poloidal and toroidal mode numbers) mode locking is easily obtained. The analysis of the mode locking thresholds varied by scanning of the spatial phase of RMPs shows that the m/n = 2/1 component of intrinsic error field of the J-TEXT tokamak is about 0.4 Gs. In addition to normal mode locking events, the (partial) stabilization of the m/n = 2 / 1 tearing mode by moderate magnetic perturbation amplitude is observed experimentally. With experimental parameters as input, both the mode locking and mode stabilization by RMPs are also obtained from nonlinear numerical modeling based on reduced MHD equations. It is found that the suppression of the tearing mode by RMPs of moderate amplitude is possible for a sufficiently high plasma rotation frequency and low Alfven velocity. Gas puffing is also used to affect the MHD activity in J-TEXT. For example, neon gas injection can cause inverse sawtooth-like activity that spreads from the q = 1 surface to the axis; in particular, small amplitude m/n = 1 / 1 mode oscillations superimposed on the inverse sawtooth waveform around the q = 1 surface are observed after the impurity injection. Nevertheless, other impurities such as helium and argon impurities can't trigger such events. In addition, gas puffing can also be applied to mitigate disruptions, especially on the current quench phase. It is found that no runaway current generation occurs in intentionally provoked disruptions when the toroidal magnetic field is lower than 2.2 T. The runaway currents can be suppressed by the intensive gas puffing of H2 . To meet the

  18. Present status of fusion research: The next step tokamak (ITER) and the demonstration reaction (DEMO)

    In the search for new sources of energy, fusion offers great possibilities for the future with virtually inexhaustible reserves and a negligible basic fuel cost. On the other hand, the conditions for the thermonuclear reactions are difficult and complex to implement because of the temperature of approximately 100 million degrees necessary to initiate nuclear combustion. Research based on magnetic confinement where magnetic field are used to contain the electrically conducting plasma was initiated in the fifties. Among the wide variety of magnetic configurations studied to date, Tokamak-type devices have obtained the best experimental results and offer the best chances of obtaining a thermonuclear plasma within the next 15 years. The european strategy for achievement of the ultimate goal of construction of a prototype electricity generating reactor has been based on three major intermediate steps:(i) the present large Tokamak JET together with other specialized devices, to prove main aspects of scientific feasibility of fusion; (ii) the planned Next Step Tokamak, to complete demonstration of scientific feasibility and to establish a solid basis for the evaluation of the technological feasibility of fusion; (iii) a DEMO reactor to complete demonstration of technological feasibility and to establish a solid basis for the evaluation of the commercial feasibility of fusion. Plasma with thermonuclear parameters have been produced in several tokamaks, JET in particular. 4 figs., 1 tab., 9 refs. (author)

  19. Results of Joint Experiments and other IAEA Activities on Research Using Small Tokamaks

    Gryaznevich, M.P.; Van Oost, G.; Peleman, P.; Brotánková, Jana; Dejarnac, Renaud; Dufková, Edita; Ďuran, Ivan; Hron, Martin; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zajac, Jaromír; Berni, L.A.; Del Bosco, E.; Ferreira, J.G.; Simões, J.R.; Berta, M.; Dunai, D.; Tál, B.; Zoletnik, S.; Malaquias, A.; Mank, G.; Figueiredo, H.; Kuznetsov, Y.; Ruchko, L.; Hegazy, H.; Ovsyannikov, A.; Sukhov, E.; Vorobjev, G.M.; Dreval, N.; Singh, A.; Budaev, V.; Kirnev, G.; Kirneva, N.; Kuteev, B.; Melnikov, A.; Nurov, D.; Sokolov, M.; Vershkov, V.; Khorshid, P.; Gonzales, R.; El Chama Neto, I.; Kraemer-Flecken, A.W.; Soldatov, V.; Brotas, B.; Carvalho, P. S.; Coelho, R.; Duarte, A.; Fernandes, H.; Figueiredo, J.; Fonseca, A.; Gomes, R.; Nedzelskiy, I.; Neto, A.; Ramos, G.; Santos, J.; Silva, C.; Valcárcel, D.; Gutierrez Tapia, C.R.; Krupnik, L.I.; Petrov, L.; Kolokoltsov, M.; Herrera, J.; Nieto-Perez, M.; Czarnecka, A.; Balan, P.; Sharnin, A.; Pavlov, V.

    Vienna : International Atomic Energy Agency, 2008, OV/P1-1-OV/P1-8. ISBN N. [IAEA Fusion Energy Conference/22nd./. Geneva (CH), 13.10.2008-18.10.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * joint experiment * turbulence * transport barrier * improvement confinement * electric field Subject RIV: BL - Plasma and Gas Discharge Physics http://www-pub.iaea.org/MTCD/Meetings/FEC2008/ov_p1-1.pdf

  20. Texas Experimental Tokamak: A plasma research facility. Technical progress report, November 1, 1993--October 31, 1994

    The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics in order to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks and in particular to understand the role of turbulence. So that they can continue to study the physics that is most relevant to the fusion program, TEXT completed a significant device upgrade this year. The new capabilities of the device and new and innovative diagnostics were exploited in all main program areas including: (1) configuration studies; (2) electron cyclotron heating physics; (3) improved confinement modes; (4) edge physics/impurity studies; (5) central turbulence and transport; and (6) transient transport. Details of the progress in each of the research areas are described

  1. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available

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

    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. Analysis of line integrated electron density using plasma position data on Korea Superconducting Tokamak Advanced Research

    A 280 GHz single-channel horizontal millimeter-wave interferometer system has been installed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This system has a triangular beam path that does not pass through the plasma axis due to geometrical constraints in the superconducting tokamak. The term line density on KSTAR has a different meaning from the line density of other tokamaks. To estimate the peak density and the mean density from the measured line density, information on the position of the plasma is needed. The information has been calculated from tangentially viewed visible images using the toroidal symmetry of the plasma. Interface definition language routines have been developed for this purpose. The calculated plasma position data correspond well to calculation results from magnetic analysis. With the position data and an estimated plasma profile, the peak density and the mean density have been obtained from the line density. From these results, changes of plasma density themselves can be separated from effects of the plasma movements, so they can give valuable information on the plasma status.

  4. Texas Experimental Tokamak, a plasma research facility: Technical progress report

    Wootton, A.J.

    1995-08-01

    In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively.

  5. Texas Experimental Tokamak, a plasma research facility: Technical progress report

    In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively

  6. Results of Joint Experiments and other IAEA activities on research using small tokamaks

    Brotánková, Jana; Dejarnac, Renaud; Dufková, Edita; Ďuran, Ivan; Hron, Martin; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zajac, Jaromír

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104026-104026. ISSN 0029-5515. [IAEA Fusion Energy Conference/22nd./. Geneva, 13.10.2008-18.10.2008] R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * probe diagnostics * sheared flows * edge plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://iopscience.iop.org/0029-5515/49/10/104026

  7. Korea Superconducting tokamak advanced research project - Development of heating system

    Choi, Byung Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The heating and current drive systems for KSTAR based on multiple technologies (neutral beam, ion cyclotron, lower hybrid and electron cyclotron) have been designed to provide heating and current drive capabilities as well as flexibility in the control of current density and pressure profiles needed to meet the mission and research objectives of the machine. They are designed to operate for long-pulse lengths of up to 300 s. The NBI system initially delivers 8 MW of neutral beam power to the plasma from one co-directed beam line and shall be upgraded to provide 20 MW of neutral beam power with two co-directed beam lines plus one counter-directed beam line. It will be capable of being reconfigured such that the source arrangement is changed from horizontal to vertical stacking, with 6 MW beam power to the plasmas per beam line, in order to facilitate profile control. The RF system initially delivers 6 MW of rf power to the plasma, using a single four-strap antenna mounted in a midplane port. The system will be upgraded to proved 12 MW of rf power through 2 adjacent ports. In the first phase, we completed the basic design of RF system and the system have the capabilities to be operationable for pulse length up to 300 sec and in the 25-60 MHz frequency range. Lower hybrid system initially provides 1.5 MW LH rf power to the plasma at 3.7 GHz through a horizontal port, which has a capability to be operated for pulse length up to 300 sec, and shall be upgraded to provide 4.5 MW of LH rf power to the plasma. In the first phase, we completed the basic design of LHCD system which incorporate the TPX-type launcher and independently phase-changeable transmission system for the fully phased coupler. The ECH system will deliver up to 0.5 MW of power to the plasma for up to 0.5 sec. In the first phase, we completed the basic design of ECH system which includes an 84 GHz gyrotron system, a transmission system, and a launcher. The basic design of the low loss transmission system

  8. Tokamak Plasmas : Internal magnetic field measurement in tokamak plasmas using a Zeeman polarimeter

    M Jagadeeshwari; J Govindarajan

    2000-11-01

    In a tokamak plasma, the poloidal magnetic field profile closely depends on the current density profile. We can deduce the internal magnetic field from the analysis of circular polarization of the spectral lines emitted by the plasma. The theory of the measurement and a detailed design of the Zeeman polarimeter constructed to measure the poloidal field profile in the ADITYA tokamak are presented. The Fabry-Perot which we have employed in our design, with photodiode arrays followed by lock-in detection of the polarization signal, allows the measurement of the fractional circular polarization. In this system He-II line with wavelength 4686 Å is adopted as the monitoring spectral line. The line emission used in the present measurement is not well localized in the plasma, necessiating the use of a spatial inversion procedure to obtain the local values of the field.

  9. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research.

    Lampert, M; Anda, G; Czopf, A; Erdei, G; Guszejnov, D; Kovácsik, Á; Pokol, G I; Réfy, D; Nam, Y U; Zoletnik, S

    2015-07-01

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera's measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties. PMID:26233377

  10. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties

  11. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    Lampert, M. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); BME NTI, Budapest (Hungary); Anda, G.; Réfy, D.; Zoletnik, S. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); Czopf, A.; Erdei, G. [Department of Atomic Physics, BME IOP, Budapest (Hungary); Guszejnov, D.; Kovácsik, Á.; Pokol, G. I. [BME NTI, Budapest (Hungary); Nam, Y. U. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-07-15

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

  12. Software upgradation of PXI based data acquisition for Aditya experiments

    Aditya Data Acquisition and Control System is designed to acquire data from diagnostics like Loop Voltage, Rogowski, Magnetic probes, X-rays etc and for control of gas feed, gate valve control, trigger pulse generation etc. CAMAC based data acquisition system was updated with PXI based Multifunction modules. The System is interfaced using optical connectivity with PC using PCI based controller module. Data is acquired using LabVIEW graphical user interface (GUI) and stored in server. The present GUI based application does not have features like module parameters configuration, analysis, webcasting etc. So a new application software using LabVIEW is being developed with features for individual module support considering programmable channel configuration - sampling rate, number of pre and post trigger samples, number of active channel selection etc. It would also have facility of using multi-functionality of timer and counter. The software would be scalable considering more modules, channels and crates along with security of different access level of user privileges. (author)

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

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

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

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

  15. Researches on the Neutral Gas Pressure in the Divertor Chamber of the HL-2A Tokamak

    WANGMingxu; LIBo; YANGZhigang; YANLongwen; HONGWenyu; YUANBaoshan; LIULi; CAOZeng; CUIChenghe; LIUYong; WANGEnyao; ZHANGNianman

    2003-01-01

    The neutral gas pressure in divertor chamber is a very basic and important physics parameter because it determines the temperature of charged particles, the thermal flux density onto divertor plates, the erosion of divertor plates, impurity retaining and exhausting, particle transportation and confinement performance of plasma in tokamaks. Therefore, the pressure measurement in divertor chamber is taken into account in many large tokamaks.

  16. First results on disruption mitigation by massive gas injection in Korea Superconducting Tokamak Advanced Research

    Massive gas injection (MGI) system was developed on Korea Superconducting Tokamak Advanced Research (KSTAR) in 2011 campaign for disruption studies. The MGI valve has a volume of 80 ml and maximum injection pressure of 50 bar, the diameter of valve orifice to vacuum vessel is 18.4 mm, the distance between MGI valve and plasma edge is ∼3.4 m. The MGI power supply employs a large capacitor of 1 mF with the maximum voltage of 3 kV, the valve can be opened in less than 0.1 ms, and the amount of MGI can be controlled by the imposed voltage. During KSTAR 2011 campaign, MGI disruptions are carried out by triggering MGI during the flat top of circular and limiter discharges with plasma current 400 kA and magnetic field 2–3.5 T, deuterium injection pressure 39.7 bar, and imposed voltage 1.1–1.4 kV. The results show that MGI could mitigate the heat load and prevent runaway electrons with proper MGI amount, and MGI penetration is deeper under higher amount of MGI or lower magnetic field. However, plasma start-up is difficult after some of D2 MGI disruptions due to the high deuterium retention and consequently strong outgassing of deuterium in next shot, special effort should be made to get successful plasma start-up after deuterium MGI under the graphite first wall.

  17. The Fusion Science Research Plan for the Major U.S. Tokamaks. Advisory report

    In summary, the community has developed a research plan for the major tokamak facilities that will produce impressive scientific benefits over the next two years. The plan is well aligned with the new mission and goals of the restructured fusion energy sciences program recommended by FEAC. Budget increases for all three facilities will allow their programs to move forward in FY 1997, increasing their rate of scientific progress. With a shutdown deadline now established, the TFTR will forego all but a few critical upgrades and maximize operation to achieve a set of high-priority scientific objectives with deuterium-tritium plasmas. The DIII-D and Alcator C-Mod facilities will still fall well short of full utilization. Increasing the run time in vii DIII-D is recommended to increase the scientific output using its existing capabilities, even if scheduled upgrades must be further delayed. An increase in the Alcator C-Mod budget is recommended, at the expense of equal and modest reductions (~1%) in the other two facilities if necessary, to develop its capabilities for the long-term and increase its near-term scientific output.

  18. The power-supply control system for the toroidal magnets of the JFT-2M nuclear fusion research tokamak

    Mitsubishi Electric has completed the control system for the flywheel-equipped DC motor-generator that powers the toroidal field coils of the JFT-2M nuclear fusion research Tokamak at the Japan Atomic Energy Research Institute. The motor-generator, which has a maximum pulse output of 2.7 kV and 19 kA, was introduced in the May '96 issue of Giho. This article reports the key features of the coil current control. Emphasis is placed on the motor-generator field control provided by the thyristor converters, which opens the contacts of the main circuit switches in addition to controlling coil current. (author)

  19. Varennes Tokamak

    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

  20. Survey of Tokamak experiments

    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)

  1. Bibliography of fusion product physics in tokamaks

    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

  2. Tokamak Plasmas : Electron temperature $(T_{e})$ measurements by Thomson scattering system

    R Rajesh; B Ramesh Kumar; S K Varshney; Manoj Kumar; Chhaya Chavda; Aruna Thakkar; N C Patel; Ajai Kumar; Aditya Team

    2000-11-01

    Thomson scattering technique based on high power laser has already proved its superoirity in measuring the electron temperature (e) and density (e) in fusion plasma devices like tokamaks. The method is a direct and unambiguous one, widely used for the localised and simultaneous measurements of the above parameters. In Thomson scattering experiment, the light scattered by the plasma electrons is used for the measurements. The plasma electron temperature is measured from the Doppler shifted scattered spectrum and density from the total scattered intensity. A single point Thomson scattering system involving a -switched ruby laser and PMTs as the detector is deployed in ADITYA tokamak to give the plasma electron parameters. The system is capable of providing the parameters e from 30 eV to 1 keV and e from 5 × 1012 cm-3-5× 1013 cm-3. The system is also able to give the parameter profile from the plasma center ( = 0 cm) to a vertical position of = +22 cm to = -14 cm, with a spatial resolution of 1 cm on shot to shot basis. This paper discusses the initial measurements of the plasma temperature from ADITYA.

  3. Planning for US ion cyclotron heating research relevant to the Compact Ignition Tokamak and Alcator C-Mod

    Ion cyclotron heating (ICH) has been chosen as the primary method for providing auxiliary heating power to the plasma in the Compact Ignition Tokamak (CIT). Sustained progress in ion cyclotron range of frequencies (ICRF) heating experiments, together with supporting technology development, continues to justify selection of this technique as the preferred one for heating CIT to ignition. However, the CIT requirements are sufficiently different from existing achievements that continued experimentation and development are needed to meet the goals of the CIT experiment with a high degree of reliability. The purpose of this report is fourfold: (1) to review briefly the physics and technology research and development (R and D) needs for ICH on CIT, (2) to review the status of and planned programs for ICH on US and international machines, (3) to propose a unified ''mainline'' R and D program specifically geared to testing components for CIT, and (4) to assess the needs for experiments including C-Mod, the Tokamak Fusion Test Reactor (TFTR), and DIII-D to provide earlier information and improved probability of success for CIT ICH. 4 refs., 4 figs., 5 tabs

  4. Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

    It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the Dα amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase

  5. Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research.

    Lee, W; Park, H K; Lee, D J; Nam, Y U; Leem, J; Kim, T K

    2016-04-01

    The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm(-1). The upper limit corresponds to the normalized wavenumber kθρe of ∼0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed. PMID:27131668

  6. Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research

    Lee, W.; Park, H. K.; Lee, D. J.; Nam, Y. U.; Leem, J.; Kim, T. K.

    2016-04-01

    The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm-1. The upper limit corresponds to the normalized wavenumber kθρe of ˜0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed.

  7. IPP Prague contributions to the IAEA technical committee meeting on research using small tokamaks

    The report contains four papers dealing with the results of the CASTOR tokamak experiments achieved during the 1992 campaign. In the first paper the results of correlation analysis of plasma density fluctuations are reported and the role of edge plasma fluctuations in the global plasma confinement is discussed. The subject of the next paper is the improved particle confinement observed in the lower hybrid current drive and in the limiter biasing experiments. In the third paper, the close connection between the magnetic fluctuations level and the intensity of hard X radiation produced by runaway electrons is pointed out. In the last paper the atomic beam source ARALLIS used for CASTOR plasma diagnostics is described and the results of its stand testing are presented. (J.U.)

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

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

  9. Summary discussion: An integrated advanced tokamak reactor

    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

  10. Texas Experimental Tokamak

    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

  11. Experimental and modeling researches of dust particles in the HL-2A tokamak

    The investigation of dust particle characteristics in fusion devices has become more and more imperative. In the HL-2A tokamak, the morphologies and compositions of dust particles are analyzed by using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDX) with mapping. The results indicate that the sizes of dust particles are in a range from 1 μm to 1 mm. Surprisingly, stainless steel spheres with a diameter of 2.5 μm–30 μm are obtained. The production mechanisms of dust particles include flaking, disintegration, agglomeration, and arcing. In addition, dynamic characteristics of the flaking dust particles are observed by a CMOS fast framing camera and simulated by a computer program. Both of the results display that the ion friction force is dominant in the toroidal direction, while the centrifugal force is crucial in the radial direction. Therefore, the visible dust particles are accelerated toriodally by the ion friction force and migrated radially by the centrifugal force. The averaged velocity of the grain is on the order of ∼ 100 m/s. These results provide an additional supplement for one of critical plasma-wall interaction (PWI) issues in the framework of the International Thermonuclear Experimental Reactor (ITER) programme. (paper)

  12. Experimental and modeling researches of dust particles in the HL-2A tokamak

    黄治辉; 严龙文; 冨田幸博; 冯震; 程钧; 洪文玉; 潘宇东; 杨青巍; 段旭如

    2015-01-01

    The investigation of dust particle characteristics in fusion devices has become more and more imperative. In the HL-2A tokamak, the morphologies and compositions of dust particles are analyzed by using a scanning electron microscopy (SEM) and an energy dispersive x-ray spectroscopy (EDX) with mapping. The results indicate that the sizes of dust particles are in a range from 1 µm to 1 mm. Surprisingly, the stainless steel spheres with a diameter of 2.5 µm–30 µm are obtained. Production mechanism of the dust particles includes flaking, disintegration, agglomeration, and arcing. In addition, dynamic characteristics of the flaking dust particles are observed by a CMOS fast framing camera and simulated by a computer program. Both of the results display that the ion friction force is dominant in the toroidal direction, while the centrifugal force is crucial in the radial direction. Therefore, the visible dust particles are accelerated toriodally by the ion friction force and migrated radially by the centrifugal force. The averaged velocity of the grain is on the order of∼100 m/s. These results provide an additional supplement for one of critical plasma-wall interaction (PWI) issues in the framework of International Thermonuclear Experimental Reactor (ITER) programme.

  13. Spherical tokamak development in Brazil

    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. The ETE spherical Tokamak project

    Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] [and others]. E-mail: ludwig@plasma.inpe.br

    1999-07-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  15. Spherical tokamak development in Brazil

    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. Tokamaks (Second Edition)

    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)

  17. The life test of a DC circuit breaker of tokamak device JT-60 for a nuclear fusion research

    In the Tokamak devices for nuclear fusion research, the construction of the current transformer circuits having plasma as the secondary circuit and the change of the primary circuit current are necessary for generating current in the plasma. This is considered to be fairly difficult in practice if conventional methods using capacitor discharge and iron core coils are employed. Considering such circumstances, it was decided for JT-60 to use an air-core current transformer coil and to employ the method of storing energy in the form of current in the coil inductance instead of a capacitor. For this reason, a DC circuit breaker is required to interrupt coil current. The authors improved an AV vacuum breaker, which had been developed as the vacuum breaker of longitudinal magnetic field type applying a magnetic field in parallel with an arc, to get the one for DC circuit for the purpose of applying it to JT-60. In this paper, the operational characteristic of the DC breaker is described, the construction and function of the life test circuit is explained, and the test results are reported. Finally, interruptions of 10,000 times at 20 kA were carried out. It is successful that the restrike of arc occurring during tens of milli-seconds after interruptions was improved to 0.05% or less for 10,000 times operations. Further, it was found that the generation of arc restrike can be reduced practically to zero with two breakers in series. (Wakatsuki, Y.)

  18. Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

    Lee, H. Y.; Hong, J. H.; Jang, J. H.; Park, J. S.; Choe, Wonho, E-mail: wchoe@kaist.ac.kr [Department of Physics, Korea Advanced Institute of Science and Technology (KAIST), 34141 Daejeon (Korea, Republic of); Impurity and Edge Plasma Research Center, KAIST, 34141 Daejeon (Korea, Republic of); Hahn, S. H.; Bak, J. G.; Lee, J. H.; Ko, W. H.; Lee, K. D.; Lee, S. H.; Lee, H. H.; Juhn, J.-W.; Kim, H. S.; Yoon, S. W.; Han, H. [National Fusion Research Institute, 34133 Daejeon (Korea, Republic of); Ghim, Y.-C. [Deparment of Nuclear and Quantum Engineering, KAIST, 34141 Daejeon (Korea, Republic of)

    2015-12-15

    It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the D{sub α} amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase.

  19. Tokamak burn control

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  20. The ARIES tokamak reactor study

    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 D3He 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

  1. The ARIES tokamak reactor study

    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{sup 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.

  2. Theory of high-beta tokamaks

    The theoretical researches on high beta tokamak are reviewed. The ballooning mode instability is thought to be the most serious problem for the high beta tokamaks, and the theoretical results on the ballooning mode instability are discussed in detail. The experimental results in high beta belt pinch devices are also discussed. (author)

  3. Engineering Design of KSTAR tokamak main structure

    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)

  4. Power and particle exhaust in tokamaks

    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.

  5. Research on High Pressure Gas Injection As a Method of Fueling, Disruption Mitigation and Plasma Termination for Future Tokamak Reactors

    2005-01-01

    High-pressure gas injection has proved to be an effective disruption mitigation technique in DⅢ-D tokamak experiments. If the method can be applied in future tokamak reactors not only for disruption mitigation but also for plasma termination and fueling, it will have an attractive advantage over the pellet and liquid injection from the viewpoint of economy and engineering design. In order to investigate the feasibility of this option, a study has been carried out with relevant parameters for conveying tubes of different geometrical sizes and for different gases.These parameters include pressure drop, lagger time after the valve's opening, gas diffusion in an ultra-high vacuum condition, and particle number contour.

  6. Fusion research in India

    The economic growth of our country demands a rapid increase in the energy output. Fusion is one such alternate clean source of energy to contribute in the energy mix towards the second half of the century, with a virtually inexhaustible fuel supply. The environmental impact of fusion would be acceptable and relatively safe. These advantages have driven the world fusion research programme since its inception. Till a pure fusion energy source is available, it is worthwhile to develop it for the benefit of conventional fission fuel preparation and other various usages. Indian National Fusion Programme was initiated by indigenously developing the first Indian Tokamak, ADITYA, successfully commissioned in 1989 and has been generating interesting scientific results on various topics. The next major program at Institute for Plasma Research (IPR) has been to construct a Steady State Superconducting Tokamak (SST-1) by mix of import and indigenous development. After successful engineering validation of the subsystems in integrated operations, successful machine operation has been continued. Since then, the machine has been upgraded with a graphite first wall. As a strategy towards leapfrogging to save time, IPR and Department of Atomic Energy (DAE) decided on India’s participation in the International Thermonuclear Experimental Reactor (ITER) as a full partner, unique features of which will be its ability to operate for long durations and at power levels ∼500 MW sufficient to demonstrate the physics of burning plasma in a power plant like environment. It will also serve as a test-bed for additional fusion power plant technologies. To accelerate the domestic fusion research programme with integration of knowledge gained from ITER, we would embark upon design of a smaller fusion machine which will use already available technologies to produce controlled fusion reactions and use it as an energetic neutron source for test of materials developed for future fusion reactors

  7. JT-60SA project for JA-EU broader approach satellite tokamak and national centralized tokamak

    JT-60 Super Advanced (JT-60SA) project is the joint project of ITER satellite tokamak by Japan and EU with Japanese Tokamak. The background, objects, device design, management of JT-60SA is stated. It consists of six chapters: the first chapter describes introduction, the second chapter states the objects of tokamak device complementing ITER, the third chapter contains research subjects and device performance such as plasma performance and demand for devices, operation scenario, control of MHD instability, and control of heat and particles, the forth chapter design of devices, the fifth chapter management and the sixth conclusion. In order to realize prototype reactor, improvement research of tokamak, development of reactor engineering technology, fusion reactor researches, tokamak theory and simulation, and social and environment safety research has to be advanced. (S.Y.)

  8. Embedded data acquisition system with MDSPlus

    Rajpal, Rachana, E-mail: rachana@ipr.res.in [Institute for Plasma Research, Gandhinagar, Gujarat (India); Patel, Jigneshkumar; Kumari, Praveena; Panchal, Vipul; Chattopadhyay, P.K.; Pujara, Harshad; Saxena, Y.C. [Institute for Plasma Research, Gandhinagar, Gujarat (India)

    2012-12-15

    This data acquisition system (DAS) is designed and developed to cater the increasing demand of Plasma Diagnostics for Aditya Tokamak as well as to support the basic physics research going on at Institute for Plasma Research. The main design criteria were to design a system with minimum resources and flexible to cater the needs of slow and fast diagnostic channels and can be easily integrated with the existing data acquisition system of Aditya Tokamak. The DAS is designed on embedded PC/104 platform. This is a multi channel system which supports standard features of commercially available DAS. The control and bus interface logic are implemented using Very High Speed Hardware Description Language (VHDL) on Complex Programmable Logic Device (CPLD). For Aditya Tokamak pulse experiment, the software application is designed such that the data is directly integrated to the MDSplus tree of Aditya DAS. The detailed hardware and software design, development and testing results will be discussed in the paper.

  9. Embedded data acquisition system with MDSPlus

    This data acquisition system (DAS) is designed and developed to cater the increasing demand of Plasma Diagnostics for Aditya Tokamak as well as to support the basic physics research going on at Institute for Plasma Research. The main design criteria were to design a system with minimum resources and flexible to cater the needs of slow and fast diagnostic channels and can be easily integrated with the existing data acquisition system of Aditya Tokamak. The DAS is designed on embedded PC/104 platform. This is a multi channel system which supports standard features of commercially available DAS. The control and bus interface logic are implemented using Very High Speed Hardware Description Language (VHDL) on Complex Programmable Logic Device (CPLD). For Aditya Tokamak pulse experiment, the software application is designed such that the data is directly integrated to the MDSplus tree of Aditya DAS. The detailed hardware and software design, development and testing results will be discussed in the paper.

  10. Tore Supra tokamak

    This part of the electricity uses chapter of the Engineers Techniques collection is entirely devoted to the technical description of Tore Supra tokamak. A thermonuclear fusion device with magnetic confinement control such as Tore Supra concentrates a huge amount of high power electro-technical and electronic equipments. These power systems play a major role and are sometimes boosted to their extreme limits. From these equipments we can find: big superconducting magnets, big cooled copper magnets, high-voltage power supplies with thyristors (320 MVA installed), several MW hyper-frequency sources, several MW accelerated atom injectors, cryogenic, heat extraction, high-vacuum pumping systems, etc.. The components developed for these applications are numerous and frequently original: superconductor for variable magnetic field, DC static circuit breaker with high switch-off capability (0.7 GVA), 2 MW tetrodes, 500 kW klystrons, 500 kW gyrotrons, very low temperature (3 deg. K) electromechanical pumps, etc.. Tore Supra is a good example of the various applications of electricity and a testimony of the constant progress of the techniques mastered by electricians. This chapter is divided in 5 parts. Part 1 gives some general informations about thermonuclear fusion research, tokamak principles and electrotechnical systems of fusion research devices. Part 2 describes the Tore Supra tokamak, its aims and specificities, its internal components, the poloidal field system and the plasma heating systems. Part 3 concerns the power pulse sources: distribution network, poloidal field power supply, plasma heating systems, and ergodic divertor power supply. Part 4 describes the permanent electric power supplies for the auxiliary systems: toroidal field, cryogenic installation, cooling-drying loops. The last chapter briefly summarizes the perspectives of nuclear fusion research. (J.S.)

  11. Termoska pro tokamak

    Řípa, Milan

    2014-01-01

    Roč. 7, prosinec (2014), s. 16-17 Institutional support: RVO:61389021 Keywords : fusion * tokamak * cryostat * ITER Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1604-termoska-pro-tokamak

  12. Edge plasma diagnostics in tokamaks

    Stöckel, Jan; Brotánková, Jana; Hron, Martin; Adámek, Jiří; Ďuran, Ivan; Van Oost, G.; Peleman, P.; Gunn, J.; Devynck, P.; Martines, E.; Schrittwieser, R.; Kocan, M.

    Kudowa Zdrój : -, 2006, s. 910-935. [Sixth International Workshop and Summer School Towards Fusion Energy - Plasma Physics, Diagnostics, Spin-offs. Kudowa Zdrój (PL), 18.09.2006-22.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * diagnostics * heating Subject RIV: BL - Plasma and Gas Discharge Physics

  13. Tokamak experimental power reactor studies

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  14. PPPL tokamak program

    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

  15. Application of neural networks and its prospect. 4. Prediction of major disruptions in tokamak plasmas, analyses of time series data

    Disruption prediction of tokamak plasma has been studied by neural network. The disruption prediction performances by neural network are estimated by the prediction success rate, false alarm rate, and time prior to disruption. The current driving type disruption is predicted by time series data, and plasma lifetime, risk of disruption and plasma stability. Some disruptions generated by density limit, impurity mixture, error magnetic field can be predicted 100 % of prediction success rate by the premonitory symptoms. The pressure driving type disruption phenomena generate some hundred micro seconds before, so that the operation limits such as βN limit of DIII-D and density limit of ADITYA were investigated. The false alarm rate was decreased by βN limit training under stable discharge. The pressure driving disruption generated with increasing plasma pressure can be predicted about 90 % by evaluating plasma stability. (S.Y.)

  16. The ETE spherical Tokamak project. IAEA report

    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)

  17. Tokamak Systems Code

    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

  18. Control of a burning tokamak plasma

    Burmeister, R.E.; Mandrekas, J.; Stacey, W.M.

    1993-03-01

    This report is a review of the literature relevant to the control of the thermonuclear burn in a tokamak plasma. Some basic tokamak phenomena are reviewed, and then control by modulation of auxiliary heating and fueling is discussed. Other possible control methods such as magnetic ripple, plasma compression, and impurity injection as well as more recent proposed methods such as divertor biasing and L- to H-mode transition are also reviewed. The applications of modern control theory to the tokamak burn control problem are presented. The control results are summarized and areas of further research are identified.

  19. Fast IR diodes thermometer for tokamak

    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

  20. Health physics around a controlled fusion research device: the Tokamak at Fontenay-aux-Roses (T.F.R.)

    The X and neutron dosimetry measurement near the magnetic confinement device for hot plasma, called T.F.R. (Tokamak, Fontenay-aux-Roses) are presented. The biological shielding consists of an ordinary concrete wall 30 cm thick; the dose rate is thus limited at 10-1 mrem per discharge (corresponding to 10 mrem per day) in the whole area frequented by people during T.F.R. operation. A numerical calculation, taking into account the true geometry and X ray reflexion by the walls and roof, and normalized to the measurements, gives some indications on the electron beam which produces X rays. The photoneutron source (up to 1010 neutrons per dischage) and the activation of the vacuum vessel result from high energy electrons (>= 10 MeV) supporting a 10 to 1,000 A current

  1. ITER tokamak device

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

    1991-07-01

    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, 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, fueling, 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., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  2. Edge plasma studies on the CASTOR tokamak

    Hron, Martin; Peleman, P.; Spolaore, M.; Martines, E.; Hronová-Bilyková, Olena; Dejarnac, Renaud; Devynck, P.; Brotánková, Jana; Sentkerestiová, Jana; Ďuran, Ivan; Gunn, J.; Stöckel, Jan; Van Oost, G.; Adámek, Jiří; van de Peppel, L.; Štěpán, Michal

    Krakow : Euratom - IPPLM Association, 2006 - (Zagorski, R.), - [IEA Large Tokamak IA Workshop on Edge Transport in Fusion plasmas. Kraków (PL), 11.09.2006-13.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * scrape-off layer * turbulence * interchange instability Subject RIV: BL - Plasma and Gas Discharge Physics http://www.etfp2006.ifpilm.waw.pl/presentations.html

  3. Analysis of deposited impurity material on the surface of the optical window of the Tokamak using LIBS

    The emission spectra emitted from the laser-induced plasma of the optical window of Aditya Tokamak have been studied to identify the eroded materials deposited on its surface. Different layers of the window, such as the impurity deposited layer, antireflection coating and main matrix of the window material, have been identified. Laser-induced breakdown spectroscopy (LIBS) spectra of the impurity layer (first layer) shows the presence of spectral lines of Fe, Cr, Ni, Mn, Mo, Cu, C and O most of which are the components of stainless steel (SS316L) used for the fabrication of the Tokamak. LIBS spectra of the antireflection coating layer (second layer) show the spectral signature of Ca and Mg, whereas in the inner layer (last layer), the spectral lines of Al, Si and B are present. The concentrations of the impurities estimated by CF-LIBS are closely related to the constituents (major and minor) of the SS316L. Principal component analysis using LIBS data was performed to differentiate the different layers (impurity, antireflection coating and main matrix) of the window. The result of the present study demonstrates the capability of LIBS as an in-situ monitoring tool for detection and quantification of elements present in the different layers of the optical window of the Tokamak. (papers)

  4. Tokamak engineering mechanics

    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.

  5. Tokamak engineering mechanics

    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

  6. Tokamak concept innovations

    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

  7. Quantify Plasma Response to Non-Axisymmetric (3D) Magnetic Fields in Tokamaks, Final Report for FES (Fusion Energy Sciences) FY2014 Joint Research Target

    Strait, E. J. [General Atomics, San Diego, CA (United States); Park, J. -K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Marmar, E. S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ahn, J. -W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Berkery, J. W. [Columbia Univ., New York, NY (United States); Burrell, K. H. [General Atomics, San Diego, CA (United States); Canik, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delgado-Aparicio, L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ferraro, N. M. [General Atomics, San Diego, CA (United States); Garofalo, A. M. [General Atomics, San Diego, CA (United States); Gates, D. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Greenwald, M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kim, K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); King, J. D. [General Atomics, San Diego, CA (United States); Lanctot, M. J. [General Atomics, San Diego, CA (United States); Lazerson, S. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Liu, Y. Q. [Culham Science Centre, Abingdon (United Kingdom). Euratom/CCFE Association; Logan, N. C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lore, J. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Menard, J. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Nazikian, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Shafer, M. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Paz-Soldan, C. [General Atomics, San Diego, CA (United States); Reiman, A. H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Rice, J. E. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Sabbagh, S. A. [Columbia Univ., New York, NY (United States); Sugiyama, L. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Turnbull, A. D. [General Atomics, San Diego, CA (United States); Volpe, F. [Columbia Univ., New York, NY (United States); Wang, Z. R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Wolfe, S. M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2014-09-30

    The goal of the 2014 Joint Research Target (JRT) has been to conduct experiments and analysis to investigate and quantify the response of tokamak plasmas to non-axisymmetric (3D) magnetic fields. Although tokamaks are conceptually axisymmetric devices, small asymmetries often result from inaccuracies in the manufacture and assembly of the magnet coils, or from nearby magnetized objects. In addition, non-axisymmetric fields may be deliberately applied for various purposes. Even at small amplitudes of order 10-4 of the main axisymmetric field, such “3D” fields can have profound impacts on the plasma performance. The effects are often detrimental (reduction of stabilizing plasma rotation, degradation of energy confinement, localized heat flux to the divertor, or excitation of instabilities) but may in some case be beneficial (maintenance of rotation, or suppression of instabilities). In general, the magnetic response of the plasma alters the 3D field, so that the magnetic field configuration within the plasma is not simply the sum of the external 3D field and the original axisymmetric field. Typically the plasma response consists of a mixture of local screening of the external field by currents induced at resonant surfaces in the plasma, and amplification of the external field by stable kink modes. Thus, validated magnetohydrodynamic (MHD) models of the plasma response to 3D fields are crucial to the interpretation of existing experiments and the prediction of plasma performance in future devices. The non-axisymmetric coil sets available at each facility allow well-controlled studies of the response to external 3D fields. The work performed in support of the 2014 Joint Research Target has included joint modeling and analysis of existing experimental data, and collaboration on new experiments designed to address the goals of the JRT. A major focus of the work was validation of numerical models through quantitative comparison to experimental data, in

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

    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. Progress and prospects in understanding the physics of tokamak experiments

    A whistle-stop tour of the diverse physics of tokamak plasma confinement. This talk will illustrate the way in which fusion research on tokamaks has led to important and interesting physics results, and discuss some of the scientific challenges still ahead before fusion's potential can be established

  10. Role of the tokamak ISTTOK on the EURATOM fusion programme

    This paper describes the role of the tokamak ISTTOK on the development of the portuguese fusion research team, in the frame of the EURATOM Fusion Programme. Main tasks on education and training, control and data acquisition, diagnostics and tokamak physics are summarized. Work carried out on ISTTOK in collaboration with foreign teams is also reported. (author)

  11. Measurement and analysis of the radiation losses in DAMAVAND Tokamak

    Radiation losses play an important role on reaching to break-even conditions in Tokamaks. In this paper the results of measurement by a bolometer in Damavand Tokamak have been presented and analyzed. Meanwhile, we have explained our future research program on the base of last modifications in the control system of the DAMAVAND.

  12. Tokamak ARC damage

    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

  13. International tokamak reactor

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  14. Cluster storage for COMPASS tokamak

    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

  15. The Thor tokamak experiment

    The main characteristics of the plasma produced in Thor tokamak discharges are described. The machine performances are outlined and the experimental results relevant to the equilibrium, the stability and the control of the discharge regimes are discussed in detail. (author)

  16. Modular tokamak magnetic system

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  17. Tokamak simulation code manual

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    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.

  18. Tokamak simulation code manual

    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

  19. The Experiments of the small Spherical Tokamak Gutta

    GUTTA is a small spherical tokamak (R = 16cm, a = 8cm, Ip = 150kA) operating at the St. Petersburg State University since 2004 in the scope of the IAEA CRP ''Joint Research using Small Tokamaks''. Main scientific activities on GUTTA include development of new and improvement of existing mathematical models of plasma control, relevant for application on large tokamaks and ITER and verification of them on GUTTA; studies on the ECRH/EBW assisted breakdown and non-solenoid plasma formation in low aspect ratio tokamak; development of diagnostics; training and education of students.In this paper design properties of Gutta will be presented. Regimes of operation of the tokamak and plasma shape parameters are described and first results of the plasma formation and start-up studied will be discussed

  20. Electron cyclotron emission diagnostics on KSTAR tokamak

    Jeong, S. H. [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Daejeon 305-353 (Korea, Republic of); Lee, K. D.; Kwon, M. [National Fusion Research Institute, 113 Gwahangno, Daejeon 305-333 (Korea, Republic of); Kogi, Y. [Fukuoka Institute of Technology, Higashiku, Fukuoka 811-0295 (Japan); Kawahata, K.; Nagayama, Y. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Mase, A. [KASTEC, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  1. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration. PMID:21033954

  2. Vesmírný tokamak na Zemi

    Řípa, Milan

    2007-01-01

    Roč. 15, č. 3 (2007), s. 12-14. ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * technology * material * tokamak * ITER Subject RIV: BL - Plasma and Gas Discharge Physics

  3. Design of selected subsystems for COMPASS tokamak operation

    Janky, Filip; Pereira, T.; Hron, Martin; Pánek, Radomír; Fernandes, H.

    Aix-en-Provence : IAEA, 2009. s. 80-80. ISBN N. [Seventh IAEATechnical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research. 15.06.2009-19.06.2009, Aix-en-Provence] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * Compass * machine control * tokamak operation Subject RIV: BL - Plasma and Gas Discharge Physics http://www-fusion-magnetique.cea.fr/tmiaea2009/ website /data/articles/000080.pdf

  4. Microwave Tokamak Experiment

    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

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

    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)

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

    Azizov, E. A.

    2012-02-01

    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.

  7. Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

    A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the 'advanced tokamak' physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs

  8. Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

    Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)

    1994-11-01

    A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.

  9. UCLA Tokamak Program Close Out Report.

    Taylor, Robert John [UCLA/retired

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  10. Material erosion and migration in tokamaks

    The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the subsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasma-facing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined

  11. First experiments with SST-1 tokamak

    Full text: SST-1, a steady state superconducting tokamak, is at advanced stage of erection at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in 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. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation and triangularity. The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 T at plasma center and a plasma current of 220 kA. Hydrogen gas will be used and plasma discharge duration will be 1000 s. 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, made up of 16 vessel sectors having ports and 16 rings with D- shaped cross-section, which are welded in-situ during the SST-1 assembly. 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. In SST-1 tokamak, the auxiliary current drive will be 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. The assembly of the SST-1 tokamak is nearing completion. The cool down of the Superconducting magnets is scheduled to start by middle of year 2004

  12. Reconnection in tokamaks

    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

  13. Advanced tokamak concepts

    Oomens, A. A. M.

    1996-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  14. Advanced tokamak concepts

    Oomens, A. A. M.

    1998-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  15. Sawtooth phenomena in tokamaks

    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

  16. IFS Numerical Laboratory Tokamak

    A numerical laboratory of a tokamak plasma is being developed. This consists of the backbone (the overall manager in terms of the MPPL programming language), and the modularized components that can be plugged in or out for a particular run and their hierarchical arrangement. The components include various metrics for overall geometry various dynamics, field calculations, and diagnoses. 2 refs

  17. Transport in gyrokinetic tokamaks

    A comprehensive study of transport in full-volume gyrokinetic (gk) simulations of ion temperature gradient driven turbulence in core tokamak plasmas is presented. Though this ''gyrokinetic tokamak'' is much simpler than experimental tokamaks, such simplicity is an asset, because a dependable nonlinear transport theory for such systems should be more attainable. Toward this end, we pursue two related lines of inquiry. (1) We study the scalings of gk tokamaks with respect to important system parameters. In contrast to real machines, the scalings of larger gk systems (a/ρs approx-gt 64) with minor radius, with current, and with a/ρs are roughly consistent with the approximate theoretical expectations for electrostatic turbulent transport which exist as yet. Smaller systems manifest quite different scalings, which aids in interpreting differing mass-scaling results in other work. (2) With the goal of developing a first-principles theory of gk transport, we use the gk data to infer the underlying transport physics. The data indicate that, of the many modes k present in the simulation, only a modest number (Nk ∼ 10) of k dominate the transport, and for each, only a handful (Np ∼ 5) of couplings to other modes p appear to be significant, implying that the essential transport physics may be described by a far simpler system than would have been expected on the basis of earlier nonlinear theory alone. Part of this analysis is the inference of the coupling coefficients Mkpq governing the nonlinear mode interactions, whose measurement from tokamak simulation data is presented here for the first time

  18. Objectives and design of the JT-60 superconducting tokamak

    A fully superconducting tokamak named as JT-60SC is designed for the modification program of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-beta steady-state operation in the use of low radio-activation ferritic steel in low ν and ρ regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented. (author)

  19. Recent experiments on the STOR-M Tokamak

    Recent experiments on the STOR-M tokamak have been focused on basic tokamak physics and technology development for controlled thermonuclear fusion research. Active control of the magnetohydrodynamic (MHD) instabilities has been achieved by helical resonant magnetic perturbations (RMPs). Improved confinement has been induced by gas puffing during ohmic discharges. Modification of toroidal flow velocities by a tangentially injected compact torus (CT) plasmoid to the STOR-M discharge has been observed. (author)

  20. The Texas Experimental Tokamak: A plasma research facility. A proposal submitted to the Department of Energy in response to Program Notice 95-10: Innovations in toroidal magnetic confinement systems

    The Fusion Research Center (FRC) at the University Texas will operate the tokamak TEXT-U and its associated systems for experimental research in basic plasma physics. While the tokamak is not innovative, the research program, diagnostics and planned experiments are. The fusion community will reap the benefits of the success in completing the upgrades (auxiliary heating, divertor, diagnostics, wall conditioning), developing diverted discharges in both double and single null configurations, exploring improved confinement regimes including a limiter H-mode, and developing unique, critical turbulence diagnostics. With these new regimes, the authors are poised to perform the sort of turbulence and transport studies for which the TEXT group has distinguished itself and for which the upgrade was intended. TEXT-U is also a facility for collaborators to perform innovative experiments and develop diagnostics before transferring them to larger machines. The general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The emphasis of the proposed research is to provide well-documented plasmas which will be used to suggest and evaluate theories, to explore control techniques, to develop advanced diagnostics and analysis techniques, and to extend current drive techniques. Up to 1 MW of electron cyclotron heating (ECH) will be used not only for heating but as a localized, perturbative tool. Areas of proposed research are: (1) core turbulence and transport; (2) edge turbulence and transport; (3) turbulence analysis; (4) improved confinement; (5) ECH physics; (6) Alfven wave current drive; and (7) diagnostic development

  1. Spontaneous generation of rotation in tokamak plasmas

    Parra Diaz, Felix [Oxford University

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  2. Radioactivity evaluation for the KSTAR tokamak

    The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 1016 s-1 through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10-4 mrem h-1 in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 106 Bq kg-1 in the carbon graphite of a plasma-facing wall. (authors)

  3. Module description of TOKAMAK equilibrium code MEUDAS

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  4. Module description of TOKAMAK equilibrium code MEUDAS

    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)

  5. Draft program plant for TNS: The Next Step after the tokamak fusion test reactor. Part III. Project specific RD and D needs

    Research and development needs for the TNS systems are described according to the following chapters: (1) tokamak system, (2) electrical power systems, (3) plasma heating systems, (4) tokamak support systems, (5) instrumentation, control, and data systems, and (6) program recommendations

  6. DIII-D research operations

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R ampersand D; and collaborative efforts

  7. Large Aspect Ratio Tokamak Study

    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

  8. Next tokamak facility

    Design studies on a superconducting, long-pulse, current-driven, ignited tokamak, called the Toroidal Fusion Core Demonstration (TFCD), are being conducted by the Fusion Engineering Design Center (FEDC) and Princeton Plasma Physics Laboratory (PPPL) with additional broad community involvement. Options include the use of all-superconducting toroidal field (TF) coils, a superconducting-copper hybrid arrangement of TF coils, or all-copper TF coils. Only the first two options have been considered to date. The general feasibility of these approaches has been established with the goal of high performance (ignition, approx. 390 MW; wall loading approx. 2.2 MW/m2) at minimum capital cost. The preconceptual effort will be completed in early FY 1984 and a selection made from the indicated options. The TFCD is judged to represent a reasonable necessary step between the Tokamak Fusion Test Reactor (TFTR) and the Engineering Test Reactor

  9. Tokamak fusion reactor exhaust

    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.)

  10. Tritium catalyzed deuterium tokamaks

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the 3He from the D(D,n)3He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general)

  11. Tokamak pump limiters

    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

  12. Microwave correllation reflectometry for tokamak CASTOR

    Nanobashvili, S.; Žáček, František; Zajac, Jaromír

    2005-01-01

    Roč. 55, č. 6 (2005), s. 701-719. ISSN 0011-4626 R&D Projects: GA AV ČR IAA1043101 Grant ostatní: GA EU(EU) INTAS ´2001 1B-2056 Institutional research plan: CEZ:AV0Z20430508 Keywords : microwaves * tokamak * plasma * turbulence * reflectometry Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.360, year: 2005

  13. Frascati Tokamak Upgrade (FTU): Results and developments

    In the present note the relation is examined between the FTU experimental programme and the most important issues in controlled thermonuclear fusion researches. FTU is a high-density, high magnetic field tokamak devoted to the study of plasma heating and current drive, energy and particle confinement and plasma-wall interaction. The most important FTU results and their relevance for ITER will be discussed

  14. Magnetic confinement experiment -- 1: Tokamaks

    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

  15. Polarization spectroscopy of tokamak plasmas

    Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs

  16. Development, calibration and performance testing of the infrared imaging video bolometer for the SST-1 Tokamak

    Infrared Imaging Video Bolometer (IRVB) is a powerful diagnostic tool for the measurement of total radiated power losses from the plasma device and it can provide temporally resolved two-dimensional (20) images of plasma radiation brightness. Recently IRVB system is designed, developed, calibrated, tested for its performance and installed on the ADITYA Tokamak for initial studies. IRVB is being developed for the first phase of SST-1 tokamak and is to be deployed at mid plane of radial port 2 with tangential viewing geometry. The IRYB developed for the SST-1 tokamak utilizes a 2.5 μm thick and 9 x 7 cm2 size free standing Platinum foil as a radiation absorber element which provides broad radiation absorptions band 1 eV to 8.5 keV (Soft X-Ray to IR). The foil is clamped on a metal frame. A pinhole camera geometry with square aperture of 0.7 x 0.7 provides 13 x 10 bolometer pixels 2-D array (130 channels) and ∼8 em of spatial resolution at the plasma mid plane with a 61° x 48° wide field of view (FOY). This wide FOY covers a tangential and a poloidal cross sectional views of SST-1 plasma. The FOY provides unique plasma viewing geometry which is confirmed by the synthetic diagnostic model results. A medium wave Infrared Camera having 320 x 240 focal plane arrays, 142 Hz full frame rate and temperature sensitivity ∼ 0.02℃ is used to record 2-D temperature distribution of the foil. Using 2-D heat diffusion analysis method, total radiated power can be estimated. The Noise Equivalent Power Density of the IRYB system has been found to be ∼ 200 μW/cm2. The present paper discusses the development and calibration of the SST-1 IRYB system. Performance of the IRVB system for its time response and NEP are experimentally investigated and has also been reported here. (author)

  17. DIII-D tokamak long range plan. Revision 3

    The DIII-D Tokamak Long Range Plan for controlled thermonuclear magnetic fusion research will be carried out with broad national and international participation. The plan covers: (1) operation of the DIII-D tokamak to conduct research experiments to address needs of the US Magnetic Fusion Program; (2) facility modifications to allow these new experiments to be conducted; and (3) collaborations with other laboratories to integrate DIII-D research into the national and international fusion programs. The period covered by this plan is 1 November 19983 through 31 October 1998

  18. Steady state operation of tokamaks. Report on the IAEA technical committee meeting held at Hefei, China, 13-15 October 1998

    The first IAEA Technical Committee Meeting on Steady State Operation of Tokamaks was held in October 1998 in Hefei, China. This meeting marks the timely start of Technical Committee Meetings in an important area of tokamak research since several experiments are already yielding impressive results and several new experiments are under construction. Among the ongoing experiments interesting results were reported from the superconducting tokamaks TRIAM 1-M, Tore Supra, and HT-7 and from a conventional tokamak, HL-1M

  19. The tokamak as a neutron source

    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

  20. Maximum entropy tokamak configurations

    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)

  1. Understanding disruptions in tokamaks

    This paper describes progress achieved since 2007 in understanding disruptions in tokamaks, when the effect of plasma current sharing with the wall was introduced into theory. As a result, the toroidal asymmetry of the plasma current measurements during vertical disruption event (VDE) on the Joint European Torus was explained. A new kind of plasma equilibria and mode coupling was introduced into theory, which can explain the duration of the external kink 1/1 mode during VDE. The paper presents first results of numerical simulations using a free boundary plasma model, relevant to disruptions.

  2. Tokamak instrumentation and controls

    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.

  3. Demonstration tokamak power plant

    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.

  4. Axisymmetric control in tokamaks

    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 sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  5. ADX - Advanced Divertor and RF Tokamak Experiment

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  6. Relativistic runaway electrons in tokamak plasmas

    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)

  7. Bifurcated Helical Core Equilibrium States in Tokamaks

    Full text: Tokamaks with weak to moderate reversed central magnetic shear in which the minimum of the inverse rotational transform qmin is in the neighbourhood of unity can trigger bifurcated MagnetoHydroDynamic (MHD) equilibrium states. In addition to the standard axisymmetric branch that can be obtained with standard Grad-Shafranov solvers, a novel branch with a three-dimensional (3D) helical core has been computed with the ANIMEC code, an anisotropic pressure extension of the VMEC code. The solutions have imposed nested magnetic flux surfaces and are similar to saturated ideal internal kink modes. The difference in energy between both possible branches is very small. Plasma elongation, current and β enhance the susceptibility for bifurcations to occur. An initial value nonlinear ideal MHD evolution of the axisymmetric branch compares favourably with the helical core equilibrium structures calculated. Peaked prescribed pressure profiles reproduce the 'snake' structures observed in many tokamaks which has led to a new explanation of the snake as a bifurcated helical equilibrium state that results from a saturated ideal internal kink in which pellets or impurities induce a hollow current profile. Snake equilibrium structures are computed in free boundary TCV tokamak simulations. Magnetic field ripple and resonant magnetic perturbations in MAST free boundary calculations do not alter the helical core deformation in a significant manner when qmin is near unity. These bifurcated solutions constitute a paradigm shift that motivates the application of tools developed for stellarator research in tokamak physics investigations. The examination of fast ion confinement in this class of equilibria is performed with the VENUS code in which a coordinate independent noncanonical phase-space Lagrangian formulation of guiding centre drift orbit theory has been implemented. (author)

  8. Tokamak exhaust process for the ITER project

    The ITER project calls for an unprecedented amount of hydrogen isotopes to be processed. To facilitate environmental responsibility and economic application of fusion technology, the re-use of hydrogen isotopes is vital. The US ITER Project Office (USIPO) is responsible for the front end of the ITER Tritium Plant, the Tokamak Exhaust Processing (TEP) System. The TEP system must separate the Tokamak exhaust gases into a stream containing only hydrogen isotopes and a stream containing only non-hydrogen gases. The USIPO has selected the Savannah River National Laboratory (SRNL) in partnership with the Los Alamos National Laboratory (LANL) to complete the TEP portion of the project. SRNL's participation builds on the laboratory's decades of work with hydrogen and its isotopes deuterium and tritium - providing the applied research and development that supports the Savannah River Site's handling of tritium. SRNL's experience and expertise in large-scale tritium processing systems and its track record of effective project execution are a unique combination that is key to the success of the ITER project. LANL brings to the partnership experience and expertise in tritium processing technologies specific to the fusion program. This knowledge and understanding were gained through the development and operation of the Tritium Systems Test Assembly at Los Alamos for over 20 years starting in the late 1970's. The US's implementation of the tokamak exhaust processing (TEP) system will provide a technically mature, robust, and cost-effective solution for the separation of hydrogen isotopes from the tokamak exhaust stream. The TEP technology, design challenges, and project status will be presented. (orig.)

  9. Application of MDSplus on EAST Tokamak

    EAST is a fully superconducting Tokamak in China used for controlled fusion research. MDSplus, a special software package for fusion research, has been used successfully as a central repository for analysed data and PCS (Plasma Control System) data since the debugging experiment in the spring of 2006. In this paper, the reasons for choosing MDSplus as the analysis database and the way to use it are presented in detail, along with the solution to the problem that part of the MDSplus library does not work in the multithread mode. The experiment showed that the data system based on MDSplus operated stably and it could provide a better performance especially for remote users

  10. Application of MDSplus on EAST Tokamak

    QU Lianzheng; LUO Jiarong; LI lingling; ZHANG Mingxing; WANG Yong

    2007-01-01

    EAST is a fully superconducting Tokamak in China used for controlled fusion research. MDSplus, a special software package for fusion research, has been used successfully as a central repository for analysed data and PCS (Plasma Control System) data since the debugging experiment in the spring of 2006 . In this paper, the reasons for choosing MDSplus as the analysis database and the way to use it are presented in detail, along with the solution to the problem that part of the MDSplus library does not work in the multithread mode. The experiment showed that the data system based on MDSplus operated stably and it could provide a better performance especially for remote users.

  11. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    Box, F.M.A.; Kolk, E. van de [Associatie Euratom-FOM, Nieuwegein (Netherlands). FOM-Instituut voor Plasmafysica; Howard, J. [Plasma Research Laboratory, Research School of Physical Science and Engineering, Australian National University, Canberra 0200 (Australia); Meijer, F.G. [Physics Faculty, University of Amsterdam, Amsterdam (Netherlands)

    1997-03-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. Several spectrometers, equipped with a charge-coupled device array, are being used with spectral ranges in the visible, the vacuum UV and the extreme UV. (orig.)

  12. Bootstrap current in a tokamak

    Kessel, C.E.

    1994-03-01

    The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and {beta}{sub p} must be kept below a critical value.

  13. Bootstrap current in a tokamak

    The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and βp must be kept below a critical value

  14. Spheromak injection into a tokamak

    Brown, M R; Bellan, P. M.

    1990-01-01

    Recent results from the Caltech spheromak injection experiment [to appear in Phys. Rev. Lett.] are reported. First, current drive by spheromak injection into the ENCORE tokamak as a result of the process of magnetic helicity injection is observed. An initial 30% increase in plasma current is observed followed by a drop by a factor of 3 because of sudden plasma cooling. Second, spheromak injection results in an increase of tokamak central density by a factor of 6. The high-current/high-density...

  15. Confinement and diffusion in tokamaks

    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*cTe/eB(δni/ni)rms which is also derived by a simple theory, the cross-field diffusion time, tp=a2/D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  16. Stellarator - tokamak configurations

    The stellarator configuration and tokamak configuration with helical fields have been studied both from an equilibrium and stability point of view. The model was restricted to a surface current model with a sharp boundary between plasma and vacuum. A general derivation of equilibrium and stability based on the Energy Principle is given. Physically the unstable modes are identified as external global modes. Detailed numerical results in different parameter regimes are presented and discussed. Critical β-limits for equilibrium and stability are obtained and in particular it is shown that in certain parameter ranges there exist a high-β as well as a low-β-region of stability. 7 refs., 14 figs

  17. New High Resolution Thomson Scattering system for the COMPASS tokamak

    Brotánková, Jana; Bělský, Petr; Weinzettl, Vladimír; Böhm, Petr

    Vol. 2. Prague : MATFYZPRESS, Prague, 2007 - (Šafránková, J.; Pavlů, J.), s. 218-223 ISBN 978-80-7378-024-1. [Annual Conference of Doctoral Students - WDS 2007 /16./. Prague (CZ), 05.06.2007-08.06.2007] R&D Projects: GA ČR GD202/03/H162 Institutional research plan: CEZ:AV0Z20430508 Keywords : Thomson Scattering * tokamak * diagnostics * laser * electron temperature * electron density * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/wds07.htm#ppm

  18. U-probe for the COMPASS Tokamak

    Kovařík, Karel; Ďuran, Ivan; Stöckel, Jan; Seidl, Jakub; Šesták, David; Brotánková, J.; Spolaore, M.; Martines, E.; Vianello, N.; Hidalgo, C.; Pedrosa, M. A.

    Prague : MATFYZPRESS, 2011 - (Šafránková, J.; Pavlů, J.), s. 227-232 ISBN 978-80-7378-185-9. - (WDS. 2). [WDS 2011 - Annual Conference of Doctoral Students /20./. Prague (CZ), 31.05.2011-03.06.2011] R&D Projects: GA ČR GD202/08/H057; GA MŠk 7G09042; GA MŠk 7G10072 Grant ostatní: EUROATOM(XE) FU07-CT-2007-00060 Institutional research plan: CEZ:AV0Z20430508 Keywords : edge plasma * filaments * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics

  19. Atomic Beam Probe Diagnostic for COMPASS Tokamak

    Háček, Pavel; Weinzettl, Vladimír; Stöckel, Jan; Anda, G.; Veres, G.; Zoletnik, S.; Berta, M.

    Vol. 2. Prague: MATFYZPRESS, 2010 - (Šafránková, J.; Pavlů, J.), s. 7-11. (WDS'10). ISBN 978-80-7378-140-8. [Annual Conference of Doctoral Students - WDS 2010 /19th./. Prague (CZ), 01.06.2010-04.06.2010] R&D Projects: GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma diagnostics * tokamak * COMPASS * beam diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics http://server.ipp.cas.cz/~vwei/work/wds2010_201_f2.pdf

  20. Progress on Joint Experiments on Small Tokamaks

    Gryaznevich, M.P.; Van Oost, G.; Del Bosco, E.; Berta, M.; Brotánková, Jana; Dejarnac, Renaud; Dufková, Edita; Ďuran, Ivan; Hron, Martin; Zajac, Jaromír; Malaquias, A.; Mank, G.; Peleman, P.; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zoletnik, S.; Tál, B.; Ferreira, J.; Fonseca, A.; Hegazy, H.; Kuznetsov, Y.; Ruchko, L.; Vorobyev, G.M.; Ovsyannikov, A.; Sukhov, E.; Singh, A.; Kuteev, B.; Melnikov, A.; Vershkov, V.; Kirneva, N.; Kirnev, G.; Budaev, V.; Sokolov, M.; Talebitaher, A.; Khorshid, P.; Ramos, G.; El Chama Neto, I.; Kraemer-Flecken, A.W.; Soldatov, V.; Marques Fonseca, A.M.; Gutierrez-Tapia, C.R.; Krupnik, L.I.

    Warsaw: EPS, 2007, P-1.070-P-1.070. (Europhysics Conference Abstracts). ISBN 978-83-926290-0-9. [EPS Conference on Plasma Physics/34th./. Warsaw (PL), 02.07.2007-06.07.2007] R&D Projects: GA AV ČR KJB100430504 Grant ostatní: EU(XE) INTAS 100008-8046 Institutional research plan: CEZ:AV0Z20430508 Source of funding: R - rámcový projekt EK Keywords : tokamak * edge plasma * turbulence * improved confinement * plasma diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics http://www.eps2007.ifpilm.waw.pl/pdf2/P1_070.pdf

  1. First Spectroscopic Measurements on the COMPASS Tokamak

    Naydenkova, Diana; Stöckel, Jan; Weinzettl, Vladimír; Šesták, David; Havlíček, Josef

    Vol. 2. Prague : MATFYZPRESS, Prague, 2009 - (Šafránková, J.; Pavlů, J.), s. 158-162 ISBN 978-80-7378-102-6. [Annual conference of doctoral students - WDS 2009 /18./. Prague (CZ), 02.06.2009-05.06.2009] R&D Projects: GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * tokamak * spectroscopic measurements Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/pdf09/WDS09_227_f2_Naydenkova.pdf

  2. Diagnostic Lithium Beam System for COMPASS Tokamak

    Háček, P.; Weinzettl, Vladimír; Stöckel, Jan; Anda, G.; Veres, G.; Zoletnik, S.; Berta, M.

    Prague : MATFYZPRESS, 2011 - (Šafránková, J.; Pavlů, J.), s. 215-220 ISBN 978-80-7378-185-9. - (WDS. 2). [WDS 2011 - Annual Conference of Doctoral Students /20./. Prague (CZ), 31.05.2011-03.06.2011] R&D Projects: GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma diagnostics * tokamak, COMPASS * beam diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics http:// server .ipp.cas.cz/~vwei/work/wds2010_201_f2.pdf

  3. Tokamak physics experiment: Diagnostic windows study

    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

  4. Electron cyclotron emission imaging in tokamak plasmas

    Munsat, Tobin; Domier, Calvin W.; Kong, Xiangyu; Liang, Tianran; Luhmann, Jr.; Neville C.; Tobias, Benjamin J.; Lee, Woochang; Park, Hyeon K.; Yun, Gunsu; Classen, Ivo. G. J.; Donne, Anthony J. H.

    2010-07-01

    We discuss the recent history and latest developments of the electron cyclotron emission imaging diagnostic technique, wherein electron temperature is measured in magnetically confined plasmas with two-dimensional spatial resolution. The key enabling technologies for this technique are the large-aperture optical systems and the linear detector arrays sensitive to millimeter-wavelength radiation. We present the status and recent progress on existing instruments as well as new systems under development for future experiments. We also discuss data analysis techniques relevant to plasma imaging diagnostics and present recent temperature fluctuation results from the tokamak experiment for technology oriented research (TEXTOR).

  5. Study of the heating of tokamaks by high energy ion beams

    This research program has encompassed a number of design studies for a steady state (or long pulse) Auto-Resonant Accelerator (ARA) capable of producing intense beams of high energy (4-20 MEV) ions suitable for the heating of large tokamak devices. The different research topics addressed have ranged over a number of questions related to the design of the individual elements of the accelerator itself, along with studies of the injection and stripping of the accelerated ions in the tokamak and their subsequent energy deposition in the tokamak plasma

  6. Measurements with an emissive probe in the CASTOR tokamak

    Schrittwieser, R.; Adámek, Jiří; Balan, P.; Hron, Martin; Ionita, C.; Jakubka, Karel; Kryška, Ladislav; Martines, E.; Stöckel, Jan; Tichý, M.; Van Oost, G.

    2002-01-01

    Roč. 44, č. 5 (2002), s. 567-578. ISSN 0741-3335 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : CASTOR tokamak, plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.121, year: 2002

  7. JOINT EXPERIMENTS ON SMALL TOKAMAKS: EDGE PLASMA STUDIES ON CASTOR

    Van Oost, G.; Berta, M.; Brotánková, Jana; Dejarnac, Renaud; Del Bosco, E.; Dufková, Edita; Ďuran, Ivan; Gryaznevich, M.P.; Horáček, Jan; Hron, Martin; Malaquias, A.; Mank, G.; Peleman, P.; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zoletnik, S.; Tál, B.; Ferrera, J.; Fonseca, A.; Hegazy, H.; Kuznetsov, Y.; Ossyannikov, A.; Singh, A.; Sokholov, M.; Talebitaher, A.

    2007-01-01

    Roč. 47, č. 5 (2007), s. 378-386. ISSN 0029-5515 R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * edge plasma * turbulence * Langmuir probe * plasma radiation * Hall probe Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.278, year: 2007

  8. Technology and physics in the Tokamak Program: The need for an integrated, steady-state RandD tokamak experiment

    The Steady-state Tokamak (STE) Experiment is a proposed superconducting-coil, hydrogen-plasma tokamak device intended to address the integrated non-nuclear issues of steady state, high-power tokamak physics and technology. Such a facility has been called for in the US program plan for the mid 1990's, and will play a unique role in the world-wide fusion effort. Information from STE on steady-state current drive, plasma control, and high power technology will contribute significantly to the operating capabilities of future steady-state devices. This paper reviews preliminary designs and expected technological contributions to the US and world fusion reactor research from each of the above mentioned reactor systems. This document is intended as a proposal and feasibility discussion and does not include exhaustive technical reviews. 12 figs., 3 tabs

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

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

  10. Moving Divertor Plates in a Tokamak

    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

  11. Fusion potential for spherical and compact tokamaks

    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

  12. Moving Divertor Plates in a Tokamak

    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.

  13. Fusion potential for spherical and compact tokamaks

    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.

  14. Tokamak experimental power reactor

    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

  15. Joint Experiments on the Tokamaks CASTOR and T-10

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confiement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. The first Joint (Host Laboratory) Experiment (JE1) has been carried out in 2005 on the CASTOR tokamak at the IPP Prague, Czech Republic. It was jointly organized by the IPP-ASCR and KFKI HAC, Budapest, involved 20 scientists from 7 countries and was supported through the IAEA and the ICTP, Trieste. The objective of JE1 was to perform studies of plasma edge turbulence and plasma confinement. Following the success of JE1, JE2 has been performed on T-10 at RRC 'Kurchatov Institute' in Moscow; 30 scientists from 13 countries participated in this experiment. This experiment aimed to continue JE1 turbulence studies, now extending them to the plasma core. Results of JE1 and JE2 will be overviewed and compared

  16. Advanced tokamak burning plasma experiment

    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)

  17. Plasma boundary phenomena in tokamaks

    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 (ne and Te) 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.)

  18. Computational studies of tokamak plasmas

    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)

  19. Controlling fusion yield in tokamaks with spin polarized fuel, and feasibility studies on the DIII-D tokamak

    The march towards electricity production through tokamaks requires the construction of new facilities and the inevitable replacement of the previous generation. There are, however, research topics that are better suited to the existing tokamaks, areas of great potential that are not sufficiently mature for implementation in high power machines, and these provide strong support for a balanced policy that includes the redirection of existing programs. Spin polarized fusion, in which the nuclei of tokamak fuel particles are spin-aligned and favorably change both the fusion cross-section and the distribution of initial velocity vectors of charged fusion products, is described here as an example of a technological and physics topic that is ripe for development in a machine such as the DIII-D tokamak. In this study, such research and development experiments may not be efficient at the ITER-scale, while the plasma performance, diagnostic access, and collaborative personnel available within the United States' magnetic fusion research program, and at the DIII-D facility in particular, provide a unique opportunity to further fusion progress

  20. Industry roles in the Tokamak Physics Experiment

    There are several distinguishing features of the Tokamak Physics Experiment (TPX) to be found in the TPX program and in the organizations for constructing and operating the machine. Programmatically, TPX addresses several issues critical to the viability of magnetic fusion power plants. Organizationally, it is a multi-institutional partnership to construct and operate the machine and carry out its program mission. An important part of the construction partnership is the integrated industrial responsibility for design, R ampersand D, and construction. The TPX physics design takes advantage of recent research on advanced tokamak operating modes achieved for time scales of the order of seconds that are consistent with continuous operation. This synergism of high performance (higher power density) modes with plasma current driven mostly by internal pressure (boot-strap effect) points toward tokamak power plants that will be cost-competitive and operate continuously. A large fraction of the project is subcontracted to industry. By policy, these contracts are at a high level in the project breakdown of work, giving contractors much of the overall responsibility for a given major system. That responsibility often includes design and R ampersand D in addition to the fabrication of the system in question. Each contract is managed through one of three national laboratories: PPPL, LLNL, and ORNL. Separate contracts for system integration and construction management round out the industry involvement in the project. This integrated, major responsibility attracts high-level corporate attention within each company, which are major corporations with long-standing interest in fusion. Through the contracts already established on the TPX project, a new standard for industry involvement in fusion has been set, and these industries will be well prepared for future fusion projects

  1. Overview on Chinese tokamak experimental progress

    Tokamak experiment research in China has made important progress. The main efforts subjected to quasi-steady state operation, LHCD, plasma heating with ICRF, IBW, NBI, ECRH, fueling with pellet and supersonic molecular beam, first wall conditioning technique. Plasma parameters in experiments were much improved, such as ne=8x1019m-3, plasma pulse >10Sec. ICRF boronization and conditioning made Zeff close to unit. Steady state full LH wave current drive has been achieved for more than 3 seconds. LHCD ramp up and recharge have also been demonstrated. The Best ηCDexp∼0.5(1+0.085 exp(4.8(BT-1.45))neICDRp/PLH=1019m-2A/W. Quasi steady state H-mode like plasma with density close to Greenwald limit was obtained by LHCD, in which energy confinement time was nearly 5 times longer than the Ohmic case. The synergy between IBW, pellet and LHCD was tested. Research on the mechanism of macro-turbulence has been extensively carried out experimentally. Ac operation of tokamak was successfully demonstrated. (author)

  2. STARFIRE: a commercial tokamak reactor

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  3. LHCD experiments on tokamak CASTOR

    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

  4. Solenoid-free plasma start-up in spherical tokamaks

    Raman, R.; Shevchenko, V. F.

    2014-10-01

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

  5. Solenoid-free plasma start-up in spherical tokamaks

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid. (topical review)

  6. RF preionization in Tokamak thor

    During the study of the RF preionization in Tokamak Thor was observed that the starting of the plasma and its time behaviour were correlated with the presence of resonance conditions both at the electron cyclotron frequency Ωsub(deg) and at its sub-harmonics Ωsub(deg)/n. These results are supported by a simple qualitative calculation

  7. Integral torque balance in tokamaks

    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.

  8. ICRF heating experiments in JFT-2 tokamak

    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)

  9. Ion cyclotron system design for KSTAR tokamak

    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.

  10. Ion cyclotron system design for KSTAR tokamak

    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

  11. Integrated plasma control for high performance tokamaks

    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. Conceptual design of Remote Control System for EAST tokamak

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication

  13. Conceptual design of Remote Control System for EAST tokamak

    Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

    2014-05-15

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

  14. Neutronics design of the next tokamak. (Swimming pool type)

    A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

  15. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  16. Transport of Dust Particles in Tokamak Devices

    Pigarov, A Y; Smirnov, R D; Krasheninnikov, S I; Rognlien, T D; Rozenberg, M

    2006-06-06

    Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.

  17. Microwave Tokamak Experiment: Overview and status

    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

  18. Module of lithium divertor for KTM tokamak

    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

    steady-state operating lithium divertor module project for Kazakhstan tokamak KTM. At present the lithium divertor module for KTM tokamak is under development in the framework of ISTC project no. K-1561. Initial heating up to 200 Degree-Sign C and lithium surface temperature stabilization during plasma interaction in the range of 350-550 Degree-Sign C will be provided by external system for thermal stabilization due to circulation of the Na-K heat transfer media. Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Development, creation and experimental research of lithium divertor model for KTM will allow to solve existing problems and to fulfill the basic approaches to designing of lithium divertor and in-vessel elements of new fusion reactor generation, to investigate plasma physics aspects of lithium influence, to develop technology of work with lithium in tokamak conditions. Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation.

  19. Bootstrap Current in Spherical Tokamaks

    王中天; 王龙

    2003-01-01

    Variational principle for the neoclassical theory has been developed by including amomentum restoring term in the electron-electron collisional operator, which gives an additionalfree parameter maximizing the heat production rate. All transport coefficients are obtained in-cluding the bootstrap current. The essential feature of the study is that the aspect ratio affects thefunction of the electron-electron collision operator through a geometrical factor. When the aspectratio approaches to unity, the fraction of circulating particles goes to zero and the contribution toparticle flux from the electron-electron collision vanishes. The resulting diffusion coefficient is inrough agreement with Hazeltine. When the aspect ratio approaches to infinity, the results are inagreement with Rosenbluth. The formalism gives the two extreme cases a connection. The theoryis particularly important for the calculation of bootstrap current in spherical tokamaks and thepresent tokamaks, in which the square root of the inverse aspect ratio, in general, is not small.

  20. Comprehensive numerical modelling of tokamaks

    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

  1. Frascati Tokamak transformer switching system

    Plasma ionization and heating, in the Frascati Tokamak, is obtained generating an emf along the plasma column, by switching the dc current flowing in the Tokamak transformer. 30 kA flowing in the 60 mH transformer inductance must be commutated into a resistance to generate 40 kV across the transformer itself. Studies and tests to solve this problem have been conducted, on different types of breakers, in cooperation between Tecnomasio Italiano Brown Boveri, Milan and Laboratori Gas Ionizzati, Frascati. Satisfactory results have finally been obtained using a DLF commercial air blast breaker in a chopper type circuit. A capacitor bank in parallel to the breaker is discharged immediately after the contacts separation and the arc in the switching element is extinguished at the first current zero. A saturable reactance in series with the breaker reduces the current decay rate to allow sufficient deionization time

  2. Burn Control Mechanisms in Tokamaks

    Hill, Maxwell; Stacey, Weston

    2013-10-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamaks, especially those used as a neutron source for fusion-fission hybrid reactors, such as the Subcritical Advanced Burner Reactor (SABR) concept. At Georgia Tech, we are developing a new burning plasma dynamics code to investigate passive safety mechanisms that could prevent power excursions in tokamak reactors. This code solves the coupled set of balance equations governing burning plasmas in conjunction with a two-point SOL-divertor model. Predictions have been benchmarked against data from DIII-D. We are examining several potential negative feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instabilities, iii) the degradation of alpha-particle confinement resulting from ripples in the toroidal field, iv) modifications to the radial current profile, v) ``divertor choking'' and vi) Type 1 ELMs.

  3. Analysis on the severe accidents in KSTAR tokamak

    Lee, Myoung Jae; Cheong, Y. H.; Choi, Y. S.; Cheon, E. J. [PlaGen, Seoul (Korea, Republic of)

    2003-11-15

    The establishment of regulatory and approval systems for KSTAR (Korea Superconducting Tokamak Advanced Research) has been demanded as the facility is targeted to be completed in the year of 2005. Such establishment can be achieved by performing adequate and in-depth analyses on safety issues covering radiological and chemical hazard materials, radiation protection, high vacuum, very low temperature, etc. The loss of coolant accidents and the loss of vacuum accident in fusion facilities have been introduced with summary of simulation results that were previously reported for ITER and JET. Computer codes that are actively used for accident simulation research are examined and their main features are briefly described. It can be stated that the safety analysis is indispensable to secure the safety of workers and individual members of the public as well as to establish the regulatory and approval systems for KSTAR tokamak.

  4. Equilibrium Reconstruction in EAST Tokamak

    QIAN Jinping; WAN Baonian; L. L. LAO; SHEN Biao; S. A. SABBAGH; SUN Youwen; LIU Dongmei; XIAO Singjia; REN Qilong; GONG Xianzu; LI Jiangang

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of toka-mak 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 ex-pansion 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.

  5. Shear Alfven waves in tokamaks

    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

  6. Instrumentation and controls of an ignited tokamak

    Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega R.J.; Raju, G.V.S.; Stone, R.S.

    1980-10-01

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented.

  7. Instrumentation and controls of an ignited tokamak

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented

  8. Magnetic confinement experiment. I: Tokamaks

    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

  9. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    Castracane, J.

    2001-01-04

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.

  10. Multi-mode remote participation on the GOLEM tokamak

    Svoboda, V.; Huang, B.; Mlynář, Jan; Pokol, G.I.; Stöckel, Jan; Vondrášek, G.

    2011-01-01

    Roč. 86, 6-8 (2011), s. 1310-1314. ISSN 0920-3796. [Symposium on Fusion Technology (SOFT) /26th./. Porto, 27.09.2010-01.10.2010] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * remote participation * education Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.490, year: 2011 http://www.sciencedirect.com/science/article/pii/S0920379611002390

  11. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies

  12. Optimization of magnetic perturbation spectra for the COMPASS tokamak

    Cahyna, Pavel

    Geneva : International Atomic Energy Agency, 2008. s. 247-247. ISBN N. [IAEA Fusion Energy Conference/22nd./. 13.10.2008-18.10.2008, Geneva] Institutional research plan: CEZ:AV0Z20430508 Keywords : resonant magnetic perturbations * ELM control * magnetic islands * saddle coils * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www-pub.iaea.org/MTCD/Meetings/PDFplus/2008/cn165/cn165_BookOfAbstracts.pdf

  13. Plasma edge biasing on CASTOR tokamak using LHCD

    Žáček, František; Petržílka, Václav; Jakubka, Karel; Stöckel, Jan; Gunn, J.; Goniche, M.; Devynck, P.; Podesta, M.; Nanobashvili, S.

    2001-01-01

    Roč. 51, č. 10 (2001), s. 1129-1138. ISSN 0011-4626. [Europhysics Workshop on Role of Electric Fields in Plasma Confinement and Exhaust/4th./. Funchal, Madeira, 24.06.2001-25.06.2001] R&D Projects: GA AV ČR IAA1043101 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.345, year: 2001

  14. KTM Tokamak is prototype of X XI century reactor. Future International laboratory of thermonuclear materials testing and power engineering

    In 29-31 May of 2000 the presentation of the joint Kazakhstan-Russian draft of Kazakhstan material-testing tokamak (KTM) was carried out. KTM tokamak is implementing by decision of the President and Government of Republic of Kazakhstan for supporting of the Kazakhstan participation in development of draft within framework of ITER fusion reactor construction. Scientific head of the project is Russian academician - Velikhov V. and Russian Research Center 'Kurchatovskij Institute' , General designers - Scientific Research Institute for Electrophysical Equipment after D. V. Efremov (Russian Federation) and Kazakh Research Inst. for Energy Industry (KazNIIEhnergoprom). Scientific part of the project is working out in National Nuclear Center of Republic of Kazakhstan and Scientific Research Institute of Experimental and Theoretical Physics. KTM tokamak is experimental fusion device for materials testing study, as well as for designing of methods for protection of the reactor first wall, in-chamber elements and divertor planes, high frequency heat of antennas in energetic load regimes close to both the ITER and the future fusion energy reactors. KTM by it design presents spheric tokamak, which successfully combining advantages of the spheromaks (compactness) and the tokamaks (high plasma density). Now in the world there are similar operating spheric tokamaks: NSTX (USA), MAST (Great Britain), GLOBUS-M (Russian Federation). Principal peculiarity of KTM tokamak is existence of moving divertor device, which with help of manipulator allows to changing of examining samples without high vacuum disruption. Values of the thermal loads and fluences in the KTM are equal or higher than loads in operating tokamaks and correspond with ITER reactor loads. KTM tokamak will be the only mega-ampere device in the world with the aspect ratio A=2

  15. Tokamak plasma position dynamics and feedback control

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form

  16. Economic evaluation of tokamak power plants

    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

  17. The disruptive instability in Tokamak plasmas

    Salzedas, F.J.B.

    2001-01-01

    Studies performed in RTP (Rijnhuizen Tokamak Project) of the most violent and dangerous instability in tokamak plasmas, the major disruption, are presented. A particular class of disruptions is analyzed, namely the density limit disruption, which occur in high density plasmas. The radiative te

  18. Physics of compact ignition tokamak designs

    Models for predicting plasma performance in compact ignition experiments are constructed on the basis of theoretical and empirical constraints and data from tokamak experiments. Emphasis is placed on finding transport and confinement models which reproduce results of both ohmically and auxiliary heated tokamak data. Illustrations of the application of the models to compact ignition designs are given

  19. The ARIES-I tokamak reactor study

    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

  20. Summary report on tokamak confinement experiments

    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

  1. Natural current profiles in a tokamak

    In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described

  2. Instrumentation for plasma diagnosis in TN (Novillo Tokamak)

    In the Plasma Physics Laboratory of National Institute of Nuclear Research it has been utilized different devices for to determine electromagnetic parameters of Novillo Tokamak such as: magnetic fields, plasma currents, plasma column position and hoop voltage. For these measurements it was designed, constructed and calibrated magnetic soundings such as: magnetic field soundings, Rogowsky coil, coils of the type called sine/cosine and spires type riding saddle; as well as the electronic instrumentation associated with these devices. This electronics to be clear of instrumentation amplifiers for the detection of the soundings signals and differentiators utilized for the elimination of spurious induced currents in the soundings by the different Novillo electromagnetic fields. In this work is presented the methodology for the construction of this instruments, as well as the results of measurements effectuated in the two operation regimens of Tokamak: Cleaning discharge and Main discharge. (Author)

  3. Design of the welded bellows for KSTAR tokamak

    Vacuum vessel of the KSTAR(Korea Superconducting Tokamak Advanced Research) tokamak is a fully welded toroidal structure with noncircular cross-section nested in the TF(Toroidal Field) coil. According to the requirements of the physics design, sixteen horizontal ports, sixteen slanted ports, sixteen baking and cooling ports, and twenty-four top and bottom vertical ports are designed for the diagnostics, plasma heating, vacuum pumping, and baking and cooling. Bellows on these ports are used for flexible components to absorb the relative displacement due to the vacuum vessel thermal expansion and the electromagnetic force between the vacuum vessel and the cryostat ports. In this study, mechanical strength of the welded bellows for KSTAR vacuum vessel was evaluated

  4. Tokamak experimental section

    Descriptions of research during this period are given for the following topics: (1) ion and electron heating, (2) high-beta and gas puff experiments, (3) beam trapping by impurities, (4) counterinjection studies, (5) impurity measurements, (6) Balmer alpha line profiles, (7) internal mode structure, (8) sawtooth oscillations and plasma transport, (9) Ormak plasma modeling, (10) charge exchange measurements, (11) wall power measurements, (12) neutron time behavior due to deuterium neutral beam injection into a hydrogen plasma, (13) wall impurities in Ormak, (14) relativistic electron studies, (15) fast x-ray energy analyzer for the 1 to 10 keV range, and (16) CTR related atomic physics

  5. Potential turbulence in tokamak plasmas

    Microscopic potential turbulence in tokamak plasmas are investigated by a multi-sample-volume heavy ion beam probe. The wavenumber/frequency spectra S(k,ω) of the plasmas potential fluctuation as well as density fluctuation are obtained for the first time. The instantaneous turbulence-driven particle flux, calculated from potential and density turbulence has oscillations of which amplitude is about 100 times larger than the steady-state outwards flux, showing sporadic behaviours. We also observed large-scale coherent potential oscillations with the frequency around 10-40 kHz. (author)

  6. The bootstrap current in tokamaks

    The properties of the Hirshman equation for the bootstrap in the tokamak and the difference between it and the simpler Hinton-Hazeltine equation are discussed. The Hirshman model, which takes into account finite-aspect-ratio effects, is used to calculate the bootstrap current in the plasma in a circular cross section with Te = Ti. Approximate upper and lower bounds on the bootstrap current are obtained. These restrict the range of variation of the current as the temperature and density profiles vary. 16 refs., 9 figs

  7. Breakdown in the pretext tokamak

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges

  8. Anomalous particle pinch in Tokamaks

    The diffusion coefficient in phase space usually varies with the particle energy. A consequence is the dependence of the fluid particle flux on the temperature gradient. If the diffusion coefficient in phase space decreases with the energy in the bulk of the thermal distribution function, the particle thermodiffusion coefficient which links the particle flux to the temperature gradient is negative. This is a possible explanation for the inward particle pinch that is observed in tokamaks. A quasilinear theory shows that such a thermodiffusion is generic for a tokamak electrostatic turbulence at low frequency. This effect adds to the particle flux associated with the radial gradient of magnetic field. This behavior is illustrated with a perturbed electric potential, for which the trajectories of charged particle guiding centers are calculated. The diffusion coefficient of particles is computed and compared to the quasilinear theory, which predicts a divergence at low velocity. It is shown that at low velocity, the actual diffusion coefficient increases, but remains lower than the quasilinear value. Nevertheless, this differential diffusion between cold and fast particles leads to an inward flux of particles. (author)

  9. Enhancement of confinement in tokamaks

    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

  10. Cluster storage for COMPASS tokamak

    Pisacka, J., E-mail: pisacka@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Hron, M., E-mail: hron@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University, V Holesovickach 2, 180 00 Praha 8 (Czech Republic); Panek, R., E-mail: panek@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer New data storage system needed for the COMPASS tokamak. Black-Right-Pointing-Pointer Distributed, fault-tolerant, parallel, scalable, non-proprietary. Black-Right-Pointing-Pointer GlusterFS selected for testing on a small test bed. Black-Right-Pointing-Pointer Aggregated reading throughput reached 300 MiB/s for 6 clients - very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  11. Cluster storage for COMPASS tokamak

    Highlights: ► New data storage system needed for the COMPASS tokamak. ► Distributed, fault-tolerant, parallel, scalable, non-proprietary. ► GlusterFS selected for testing on a small test bed. ► Aggregated reading throughput reached 300 MiB/s for 6 clients – very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  12. Predictive Modeling of Tokamak Configurations*

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

  13. Tokamak Physics Experiment divertor design

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m2. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services

  14. Atomic physics in tokamak plasmas

    Tokamak discharges produce hydrogen-isotope plasmas in a quasi-steady state, with radial electron temperature, Tsub(e)(r), and density nsub(e)(r), distribution usually centrally peaked, with typical values Tsub(e)(0) approx.= 1 - 3 keV, nsub(e)(r) approx.= 1014 cm-3. Besides hydrogen, the plasma contains small quantities of carbon, oxygen, various construction or wall-conditioning materials such as Fe, Cr, Ni, Ti, Zr, Mo, and perhaps elements added for special diagnostic purposes, e.g., Si, Sc, Al, or noble gases. These elements are spatially fairly homogeneously distributed, with the different ionization states occurring near radial locations where Tsub(e)(r) approx.= Esub(i), the ionization potential. Thus, spectroscopic measurements of various plasma properties, such as ion temperatures, plasma motions or oscillations, radial transport rates, etc. are automatically endowed with spatial resolution. Furthermore the emitted spectra, even of heavier elements such as Fe or Ni, are fairly simple because only the ground levels are appreciably populated under the prevailing plasma conditions. Identification of near-ground transitions, including particularly magnetic dipole and intercombination transitions of ions with ionization potentials in the several keV range, and determination of their collisional and radiative transition probabilities will be required for development of appropriate diagnostics of tokamak-type plasma approaching the prospective fusion reactor conditions. (orig.)

  15. Tokamak equilibria with nearly zero central current: the current hole

    The observation of stable sustainment of the 'current hole', namely the nearly zero current density region in the central part of a tokamak plasma, has opened a new class of configurations in tokamak plasmas, and a variety of research from the viewpoints of equilibrium, magnetohydrodynamics (MHD) stability, particle orbits and radial transport has been generated. Some theories and codes have been tested and extended by being applied to extreme conditions in the current hole with very weak poloidal field. The current hole is generated due to a transient negative toroidal electric field established when a large off-axis non-inductive current is rapidly formed. It has been observed in high confinement plasmas with a large fraction of bootstrap current in advanced tokamak operation. The current hole is very stiff against current drive, which suggests that it is a saturated or self-organized system. Appearance of the current hole in ITER and DEMO would be expected in some of the operation scenarios, and its influence and its control methods have been studied. Results of experimental and theoretical studies on the current hole are reviewed. (review article)

  16. A quasi-linear gyrokinetic transport model for tokamak plasmas

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

  17. Analysis of neutral hydrogenic emission spectra in a tokamak

    Balmer-α radiation by the excitation of thermal and fast neutral hydrogenic particles has been investigated in a magnetically confined fusion device, or tokamak, from the Korea Superconducting Tokamak Advanced Research (KSTAR). From the diagnostic point of view, the emission from thermal neutrals is associated with passive spectroscopy and that from energetic neutrals that are usually injected from the outside of the tokamak to the active spectroscopy. The passive spectroscopic measurement for the thermal Balmer-α emission from deuterium and hydrogen estimates the relative concentration of hydrogen in a deuterium-fueled plasma and therefore, makes a useful tool to monitor the vacuum wall condition. The ratio of hydrogen to deuterium obtained from this measurement qualitatively correlates with the energy confinement of the plasma. The Doppler-shifted Balmer-α components from the fast neutrals features the spectrum of the motional Stark effect (MSE) which is an essential principle for the measurement of the magnetic pitch angle profile. Characterization of this active MSE spectra, especially with multiple neutral beam lines crossing along the observation line of sight, has been done for the guideline of the multi-ion-source heating beam operation and for the optimization of the narrow bandpass filters that are required for the polarimeter-based MSE diagnostic system under construction at KSTAR

  18. Interactive exploration of tokamak turbulence simulations in virtual reality

    We have developed an immersive visualization system designed for interactive data exploration as an integral part of our computing environment for studying tokamak turbulence. This system of codes can reproduce the results of simulations visually for scrutiny in real time, interactively and with more realism than ever before. At peak performance, the VR system can present for view some 400 coordinated images per second. The long term vision this approach targets is a open-quote holodeck-like close-quote virtual-reality environment in which one can explore gyrofluid or gyrokinetic plasma simulations interactively and in real time, visually, with concurrent simulations of experimental diagnostic devices. In principle, such a open-quote virtual tokamak close-quote computed environment could be as all encompassing or as focussed as one likes, in terms of the physics involved. The computing framework in one within which a group of researchers can work together to produce a real and identifiable product with easy access to all contributions. This could be our version of NASA's next generation Numerical Wind Tunnel. The principal purpose of this VR capability for Numerical Tokamak simulation is to provide interactive visual experience to help create new ways of understanding aspects of the convective transport processes operating in tokamak fusion experiments. The effectiveness of the visualization method is strongly dependent on the density of frame-to-frame correlation. Below a threshold of this quantity, short term visual memory does not bridge the gap between frames well enough for there to exist a strong visual connection. Above the threshold, evolving structures appear clearly. The visualizations show the 3D structure of vortex evolution and the gyrofluid motion associated with it. We discovered that it was very helpful for visualizing the cross field flows to compress the virtual world in the toroidal angle

  19. Interlock system for the COMPASS tokamak

    Hron, Martin; Sova, J.; Šíba, J.; Kovář, J.; Adámek, Jiří; Pánek, Radomír; Havlíček, Josef; Písačka, Jan; Mlynář, Jan; Stöckel, Jan

    2010-01-01

    Roč. 85, 3-4 (2010), s. 505-508. ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition and Remote Participation for Fusion Research/7th./. Aix – en – Provence, 15.06.2009-19.06.2009] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak operation * Interlock * Personnel safety Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.143, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V3C-5003BXW-1&_user=6542793&_coverDate=07%2F31%2F2010&_rdoc=1&_fmt=high&_orig=search&_origin=search&_sort=d&_docanchor=&view=c&_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=ef5794d05cc6530a905d1de43aa0ac6a&searchtype=a

  20. Plasma equilibrium and instabilities in tokamaks

    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.)

  1. Robust Sliding Mode Control for Tokamaks

    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.

  2. The ARIES-I tokamak reactor study

    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

  3. Synchrotron radiation in inhomogeneous tokamak plasmas

    Synchrotron emission in a tokamak configuration with inhomogeneous plasma parameters is considered to investigate the effects of the temperature profile and vertical elongation on the radiation loss. Using the numerical solution of the transfer equation for ITER-like plasma parameters, several new results on the radiated energy in a Maxwellian plasma have been derived. In particular: (i) synchrotron loss is profile dependent, namely, at constant average thermal energy, the emitted radiation increases with the peak temperature, (ii) an analytical formula of the global loss in inhomogeneous tokamak plasmas with arbitrary vertical elongation is established, (iii) the maximum of the frequency emission spectrum is a linear function of the volume average temperature, (iiii) high frequency synchrotron radiation is entirely due to electrons with energy much greater than the thermal energy. The need for experimental investigations on synchrotron emission in present-day large tokamaks to determine the effect of reflections of the complex tokamak first wall is stressed

  4. D-D tokamak reactor studies

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated

  5. Plasma diagnostics using synchrotron radiation in tokamaks

    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.

  6. Thermonuclear ignition in the next generation tokamaks

    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 aBtx 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

  7. Epoxide insulation for Tokamak coils

    The construction and testing of 12-tonne toroidal-field electromagnets for the Joint European Torus by Brown Boveri and Cie (Mannheim) are described. The principle of Tokamak confinement of a plasma which acts as the secondary winding of a transformer is explained. The Cu conductors are sanded and coated with epoxide adhesive before being wrapped in 7mm thick woven glass fibre, dried by heating under vacuum, impregnated and encapsulated in 1.2 tonnes of Araldite, which is solidified under pressure of 4 atmospheres and hardened for ten hours at 1500C. The prototype withstood tests involving 25,000 flexure cycles at 1.1 MN and 2 Hz, 2,000 quarter-hour 10kA heating cycles between 840 and 200C, and exposure to 500 million rads. 32 such coils were constructed at the rate of one every three weeks. (M.B.D.)

  8. Tokamak plasma interaction with limiters

    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 D2 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

  9. Study of electron density and its fluctuations in tokamaks plasmas by fast infrared interferometry

    The electron density knowledge in tokamak plasma is fundamental for controlled fusion research. Its study can be made by interferometric measurement of plasma refraction index. Density and density fluctuation measurements are given for present and future tokamak, the wavelength used must be in the far infrared. The interferometer used type employs two identical lasers. Waveguide type submillimetric lasers, optically pumped by a CO2 laser, have been developed and optimized. Detectors used are Schottky diodes. The interferometer allows a radial study of the plasma and presents a great stability during the measurement

  10. Applications of non-resonant RF forces for improvement of tokamak reactor performances, 1

    Applications of Radio Frequency to a tokamak divertor plasma for improvement of its reactor relevancy is studied and proposed. RF is applied to a divertor region of a tokamak by use of wave guide launchers on consideration of a reactor environment. Since the ponderomotive force is dependent on the charge to mass ratio of ions, various useful applications are considered. They covers some of the recent key issues in the development research toward nuclear fusion: the reduction of the heat load on the divertor plate, improvement of tritium inventory, impurity control, and helium ash removal. (author)

  11. Importance of the fine structure in a tokamak for the abnormal transport and the internal disruptions

    The problem of energy transport in a Tokamak, in presence of magnetic islets, has been treated by decomposing this problem in different bricks. To assembly the different bricks the model of dynamic percolation, which couples by the intermediate of scattering coefficient, the activity of transport sites (islets size) to the profile of transported quantity (temperature profile) has been chosen. The results, got with this model, results connected to the hypothesis of a limited number of islets, agree with the different observations. A possible application of this model could be the exploration of different operating conditions of Tokamak and a research of improved confinement running. (N.C.). 149 refs., 85 figs

  12. Annual progress report on fusion plasma theory task III: auxiliary heating in tokamaks and tandem mirrors

    The research we have accomplished during the past year has focussed on ICRF coupling, heating and breakeven studies for tokamaks and ECRF fundamental second harmonic heating in tandem mirrors. The studies have included ICRF Fokker-Planck heating and breakeven studies for large tokamaks such as JET, fundamental work on a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas, a formulation and code development for ICRF waveguide coupling in tokamak edge regions. ECRF ray tracing studies have been carried out for fundamental and second harmonic propagation, absorption and whistler microinstabilities in tandem mirror plug and barrier regions of Phaedrus, TMX-U and TASKA. The two-dimensional velocity space, time dependent Fokker-Planck heating studies have concentrated on D-T breakeven scenarios for fundamental minority deuterium and second harmonic tritium regimes

  13. On Use of Semiconductor Detector Arrays on COMPASS Tokamak

    Weinzettl, Vladimír; Imríšek, Martin; Havlíček, Josef; Mlynář, Jan; Naydenkova, Diana; Háček, Pavel; Hron, Martin; Janky, Filip; Sarychev, D.; Berta, M.; Bencze, A.; Szabolics, T.

    -, č. 71 (2012), s. 844-850. ISSN 2010-376X. [ICPP 2012 : International Conference on Plasma Physics. Venice, 14.11.2012-16.11.2012] R&D Projects: GA ČR GA202/09/1467; GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : bolometry * plasma diagnostics * soft X-rays * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics https://www.waset.org/journals/waset/v71/v71-143.pdf

  14. A DESIGN RETROSPECTIVE OF THE DIII-D TOKAMAK

    OAK-B135 The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and rf heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research program. This paper gives an integrated picture of the facility and its capabilities

  15. A design retrospective of the DIII-D tokamak

    Luxon, J. L.

    2002-05-01

    The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and RF heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research programme. An integrated picture of the facility and its capabilities is presented.

  16. EBW H&CD Potential for Spherical Tokamaks

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

    Vol. 1406. New York : American Institute of Physics, 2011 - (Phillips, C.; Wilson, J.), s. 477-480 ISBN 978-0-7354-0978-1. - (AIP Conference Proceedings. 1406). [Topical Conference on Radio Frequency Power in Plasmas/19./. Newport (US), 01.06.2011-03.06.2011] R&D Projects: GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : Fusion * tokamak * heating * current drive * electron Bernstein wave * EBW Subject RIV: BL - Plasma and Gas Discharge Physics http://scitation.aip.org/getpdf/servlet/GetPDFServlet?filetype=pdf&id=APCPCS001406000001000477000001&idtype=cvips&doi=10.1063/1.3665018&prog=normal

  17. Vacuum Control and Gas Handling for COMPASS Tokamak

    Janky, F.; Pereira, T.V.; Santos, B.A.; Hron, Martin

    Vol. 2. Prague: MATFYZPRESS, Prague, 2009 - (Šafránková, J.; Pavlů, J.), s. 153-157 ISBN 978-80-7378-102-6. [Annual conference of doctoral students - WDS 2009 /18./. Prague (CZ), 02.06.2009-05.06.2009] R&D Projects: GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : vacuum * control * tokamak COMPASS * gas handling Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/pdf09/WDS09_226_f2_Janky.pdf

  18. Computer and engineering calculations of Brazilian Tokamak-II

    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.)

  19. First dedicated observations of runaway electrons in the COMPASS tokamak

    Vlainić Miloš

    2015-06-01

    Full Text Available Runaway electrons present an important part of the present efforts in nuclear fusion research with respect to the potential damage of the in-vessel components. The COMPASS tokamak a suitable tool for the studies of runaway electrons, due to its relatively low vacuum safety constraints, high experimental flexibility and the possibility of reaching the H-mode D-shaped plasmas. In this work, results from the first experimental COMPASS campaign dedicated to runaway electrons are presented and discussed in preliminary way. In particular, the first observation of synchrotron radiation and rather interesting raw magnetic data are shown.

  20. Evaluation of the plasma parameters in COMPASS tokamak divertor area

    Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud

    2012-01-01

    Roč. 356, č. 1 (2012), s. 012007. ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf

  1. Motion of Charged Particles in Perturbed Magnetic Fields of Tokamak

    Papřok, R.; Krlín, Ladislav; Cahyna, Pavel; Riccardo, V.

    Vol. 2. Prague: MATFYZPRESS, Prague, 2009 - (Šafránková, J.; Pavlů, J.), s. 139-143 ISBN 978-80-7378-102-6. [Annual conference of doctoral students - WDS 2009 /18./. Prague (CZ), 02.06.2009-05.06.2009] Institutional research plan: CEZ:AV0Z20430508 Keywords : resonant magnetic perturbations * magnetic islands * electric field * runaway electrons * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/contents/pdf09/WDS09_223_f2_Paprok.pdf

  2. First measurements with U-probe on the COMPASS tokamak

    Kovařík, Karel; Ďuran, Ivan; Stöckel, Jan; Seidl, Jakub; Šesták, David; Brotánková, J.; Spolaore, M.; Martines, E.; Vianello, N.

    Vol. 2. Prague : MATFYZPRESS, 2013 - (Šafránková, J.; Pavlů, J.), s. 109-114 ISBN 978-80-7378-251-1. - (WDS). [Annual Conference of Doctoral Students – WDS 2013 /22./. Praha (CZ), 04.06.2013-07.06.2013] R&D Projects: GA MŠk 7G09042; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/proc/pdf13/WDS13_217_f2_Kovarik.pdf

  3. Focus on nuclear fusion research

    Křenek, Petr; Mlynář, Jan

    2011-01-01

    Roč. 61, - (2011), s. 62-63. ISSN 0375-8842 Institutional research plan: CEZ:AV0Z20430508 Keywords : ITER * COMPASS * fusion energy * tokamak * EURATOM Subject RIV: BL - Plasma and Gas Discharge Physics http://www.ipp.cas.cz/Tokamak/clanky/energetika_COMPASS.pdf

  4. Particle and energy balances in tokamak plasmas

    Computational and experimental studies on particle and energy balances in tokamak plasmas are described. Firstly, concerning the modeling of tokamak plasmas, the particle balance considering diffusion and recycling, and the energy balance considering transport and energy losses due to impurities are discussed. Production mechanisms of gaseous and metallic impurities, which play important role in tokamak plasmas, are also discussed from a viewpoint of plasma-wall interactions. Scaling laws of density, temperature and energy confinement time are shown on the basis of recent data. Secondarily, tokamak plasmas are simulated with the above model, and anomalous diffusion and electron thermal conduction are indicated. Characteristics of a future tokamak plasma are also simulated. Stationary impurity density distributions and related energy losses, such as bremsstrahlung, ionization and excitation, are calculated taking into account diffusion and ionization processes. Edge cooling by oxygen impurities is described quantitatively compared with experiments. Permissible impurity levels of carbon, oxygen and iron in future large tokamaks are estimated. Thirdly, experimental studies on surface cleaning methods of the first wall are described; discharge cleaning in JFT-2, baking effect on the outgassing rates of wall materials, surface treatment of high-temperature molybdenum by oxygen and hydrogen gases, and in-situ coating of molybdenum by a coaxial magnetron sputter method. Lastly, problems in future large tokamaks aiming at break-even or self-ignited plasma are discussed quantitatively, such as trapped particle instabilities, impurities and additional heating. It is predicted that new conceptions will be necessary to overcome the problems and attain the fusion goal. (auth.)

  5. Short-term power sources for tokamaks and other physical experiments

    Zajac, Jaromír; Žáček, František; Brettschneider, Zbyněk; Lejsek, V.

    2007-01-01

    Roč. 82, č. 4 (2007), s. 369-379. ISSN 0920-3796 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * Impulse power sources * Energy accumulation Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 1.058, year: 2007 http://www.sciencedirect.com/science/journal/09203796

  6. Tokamak edge electron diffusion and distribution function in the lower hybrid antenna electric field

    Fuchs, Vladimír; Gunn, J. P.; Goniche, M.; Petržílka, Václav

    2003-01-01

    Roč. 43, č. 5 (2003), s. 341-351. ISSN 0029-5515 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, grill electric field Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.390, year: 2003

  7. Emissive probe measurements of plasma potential fluctuations in the edge plasma regions of tokamaks

    Balan, P.; Schrittweiser, R.; Ionita, C.; Cabral, J. A.; Figueiredo, H. F. C.; Fernandes, H.; Varandas, C.; Adámek, Jiří; Hron, Martin; Stöckel, Jan; Martines, E.; Tichý, M.; Van Oost, G.

    2003-01-01

    Roč. 74, č. 3 (2003), s. 1583-1587. ISSN 0034-6748 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : plasma physics, tokamaks, probes Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.343, year: 2003

  8. Measurement of the Fluctuation-Induced Flux with Emissive Probe in the CASTOR Tokamak

    Balan, P.; Adámek, Jiří; Ďuran, Ivan; Hron, Martin; Ionita, C.; Martines, E.; Schrittwieser, R.; Stöckel, Jan; Tichý, M.; Van Oost, G.

    Mulhouse: European Physical Society, 2002 - (Behn, R.; Varandas, C.), s. P-2.072 [EPS Conference on Controlled Fusion and Plasma Physics /29./. Montreux (CH), 17.06.2002-21.06.2002] Institutional research plan: CEZ:AV0Z2043910 Keywords : CASTOR tokamak, electric field Subject RIV: BL - Plasma and Gas Discharge Physics

  9. Role of turbulence and electric fields in the establishment of improved confinement in tokamak plasmas

    Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Krämer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Peleman, P.; Razumova, K.; Stöckel, Jan; Vershkov, V.; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Dnestrovskij, A.Yu.; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Jachmin, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Soldatov, S.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Zimmermann, O.

    2006-01-01

    Roč. 12, č. 6 (2006), s. 14-19. ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * improved confinement * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics http://vant.kipt.kharkov.ua/TABFRAME.html

  10. The JT-60 tokamak machine

    JT-60 is a large tokamak experimental device under construction at JAERI with main device parameters of R=3.0m, a=0.95m, Bsub(t)=45kG, and Isub(p)=2.7Ma. Its basic aim is to produce and confine hydrogen plasmas of temperatures in a multi-keV range and of confinement times comparable to a second, and to study its plasma-physics properties as well as engineering problems associated with them. The JT-60 tokamak machine is mainly composed of a vacuum vessel, toroidal field (TF) coils, poloidal field (PF) coils, and support structures. The vacuum vessel is a high toroidal chamber with an egg-shaped crossection, consisting of sectorial rigid rings and parallel bellows made from Inconel 625. It is baked out at a maximum temperature up to 5000C. Several kinds of first walls made from molybdenum are bolt-jointed to the vacuum vessel for its protection. The vacuum vessel is almost completely finished with design and is deeply into manufacturing. The TF system consists of 18 unit coils located around a torus axis at regular intervals. The unit coil composed of two pancakes are wedge-shaped at the section close to a torus axis and encased in a high-manganese non-magnetic steel case. Fabrication of the TF coils will be finished in May 1981. The PF coils are composed of ohmic heating coils, vertical field coils, horizontal field coils, and quadrupole field coils located inside the TF coil bore and outside the vacuum vessel, and magnetic limiter coils placed in the vacuum vessel. Its mechanical and thermal design is almost completed are composed of the upper and lower support structures, support comuns of the vacuum vessel, and central column made from high-manganese non-magnetic steel. The structural analysis was completed including a seismic analysis and the fabrication is now in progress. The first plasma is expected to be produced in October 1984. (orig.)

  11. Heavy Neutral Beam Probe for edge plasma analysis in Tokamaks

    The contents of this report present the progress achieved to date on the Heavy Neutral Beam Probe project. This effort is an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadien de Fusion Magnetique (CCFM). The overall objective of the effort is to develop and apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes (TdeV) facility in Montreal, Canada. To achieve this goal, a research and development project was established to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present the project is in the middle of its second budget period with the instrumentation on-site at TdeV. The first half of this budget period was used to complete total system tests at InterScience, Inc., dismantle and ship the hardware to TdeV, re-assemble and install the HNBP on the tokamak. Integration of the diagnostic into the TdeV facility has progressed to the point of first beam production and measurement on the plasma. At this time, the HNBP system is undergoing final de-bugging prior to re-start of machine operation in early Fall of this year

  12. Plasma Physics Regimes in Tokamaks with Li Walls

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors

  13. Small tokamaks for fusion technology testing

    Small steady-state tokamaks for testing divertors and fusion nuclear technologies are considered. Based on present physics and technology data and explanation to reduce R0/a, H-D-fueled tokamaks with R0 ∼ 0.6--0.75 m, R0/a ∼ 1.8--2.5, and Bt0 ∼ 1.4--2.2 T can be driven with Ptot ∼ 4.5 MW to maintain Ip ∼ 0.5 MA and produce the ITER-level plasma edge and divertor conditions. Given an adequate steady-state divertor solution and Q∼1 operation based on fusion through the suprathermal component, D-T-fueled tokamaks with R0 ∼ 0.8 m, R0/a ∼ 2, and Bt0 ∼ 4 T can be driven with Ptot ∼ 15 MW to maintain Ip ∼ 4.6 MA and produce an peak neutron wall load WL ∼ 1 MW/m2. Such devices appear possible if the plasma properties at the power R0/a remain tokamak-like and, for the D-T case, can unshielded center core is feasible. The use of a single conductor as the inboard leg of the toroidal field coils for this purpose is discussed. The physics issues and the design features are identified for such tokamaks with a testing duty for factor goal of 10--20%

  14. Three novel tokamak plasma regimes in TFTR

    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

  15. Three novel tokamak plasma regimes in TFTR

    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.

  16. Electron thermal transport in tokamak plasmas

    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 (108 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)

  17. Simulation of burning tokamak plasmas

    To simulate dynamical behaviour of tokamak fusion reactors, a zero-dimensional time-dependent particle and power balance code has been developed. The zero-dimensional plasma model is based on particle and power balance equations that have been integrated over the plasma volume using prescribed profiles for plasma parameters. Therefore, the zero-dimensional model describes the global dynamics of a fusion reactor. The zero-dimensional model has been applied to study reactor start-up, and plasma responses to changes in the plasma confinement, fuelling rate, and impurity concentration, as well as to study burn control via fuelling modulation. Predictions from the zero-dimensional code have been compared with experimental data and with transport calculations of a higher dimensionality. In all cases, a good agreement was found. The advantage of the zero-dimensional code, as compared to higher-dimensional transport codes, is the possibility to quickly scan the interdependencies between reactor parameters. (88 refs., 58 figs., 6 tabs.)

  18. THOR tokamak magnetic field system

    The THOR Machine is an iron cored Tokamak having a major radius of 0.52 m and a minor radius of 0.17 m giving an aspect ratio of 3:1. It has a low ripple toroidal field of 1 T and an iron core giving 0.24 Vs. The maximum plasma current is expected to be in the region of 80x103 A. The maximum toroidal field ripple on axis is of the order of 0.01% and 2.5% at the plasma edge. The equilibrium of the plasma is achieved by means of a D.C. vertical field and a 1 cm thick copper shell. The D.C. field is cancelled during the rise time of the plasma current by means of pulsed reverse vertical field windings placed between the copper shell and the vacuum vessel. The design of this field system represents a compromise between obtaining adequate field penetration through the relatively thin vacuum vessel and maintaining the mechanical strength necessary to withstand the transient magnetic forces. Energy for the toroidal field system is supplied by a 15 kV 600 kJ capacitor bank and for the ohmic heating and reverse vertical fields by 5 kV 25 kJ and 50 kJ banks respectively. The problems encountered in the design, development and manufacture of these field systems are discussed. (author)

  19. Stability analysis of tokamak plasmas

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  20. Microtearing modes in tokamak discharges

    Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.

    2016-06-01

    Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.

  1. Tokamak x ray diagnostic instrumentation

    Three classes of x-ray diagnostic instruments enable measurement of a variety of tokamak physics parameters from different features of the x-ray emission spectrum. (1) The soft x-ray (1 to 50 keV) pulse-height-analysis (PHA) diagnostic measures impurity concentrations from characteristic line intensities and the continuum enhancement, and measures the electron temperature from the continuum slope. (2) The Bragg x-ray crystal spectrometer (XCS) measures the ion temperature and neutral-beam-induced toroidal rotation velocity from the Doppler broadening and wavelength shift, respectively, of spectral lines of medium-Z impurity ions. Impurity charge state distributions, precise wavelengths, and inner-shell excitation and recombination rates can also be studied. X rays are diffracted and focused by a bent crystal onto a position-sensitive detector. The spectral resolving power E/ΔE is greater than 104 and time resolution is 10 ms. (3) The x-ray imaging system (XIS) measures the spatial structure of rapid fluctuations (0.1 to 100 kHZ) providing information on MHD phenomena, impurity transport rates, toroidal rotation velocity, plasma position, and the electron temperature profile. It uses an array of silicon surface-barrier diodes which view different chords of the plasma through a common slot aperture and operate in current (as opposed to counting) mode. The effectiveness of shields to protect detectors from fusion-neutron radiation effects has been studied both theoretically and experimentally

  2. Measurement 20-200 keV hard X-ray based on CdTe detector in EAST Tokamak

    Background: Accurate and quantitative measurement of plasma radiation is a key issue to Tokamak, toroidal magnetic confinement device. The radiations from Tokamak cover large energy range. Driven by the determination of the obtaining of hard X-ray spectra, a new system based on a high performance CdTe detector was built up in EAST Tokamak, the first non-circle cross-section in the world. Purpose: Introduces the device of hard X-ray diagnosis system in the EAST Tokamak on the Port A. The system can measure the plasma hard X-ray (20-200 keV) spectra under different discharge conditions, including Ohmical shot and Lower Hybrid Current Drive (LHCD) shot. The research of high speed electron which produced by LHCD is also the aim of the new system. Methods: A high performance CdTe detector was using in EAST Tokamak to measure the hard X-ray (20-200 keV) spectra. Results: The results show that the new system based on a high performance CdTe can meet the requirements for measuring the EAST Tokamak. Conclusions: A preliminary experimental result showed that the system can meet the requirements for measuring the X-ray bremsstrahlung of plasma in the energy range from 20 to 200 keV Calibration result and typical measurement result on EAST are present in this paper. (authors)

  3. Extremely shaped plasmas to improve the Tokamak concept

    experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is

  4. Simulation of runaway electrons in Tokamak disruptions

    Self-consistent modelling of the generation of runaway electrons and the evolution of the toroidal electric field during tokamak disruptions is presented. The process of runaway generation is analysed by combining a relativistic kinetic equation for the electrons with Maxwell's equations for the electric field. Such modelling allows for a quantitative assessment of the runaway generation during disruptions in present day tokamak experiments, and to extrapolate to future tokamaks like ITER. It is found that the current profile can change dramatically during a disruption, such that the post disruption current, carried mainly by the runaway electrons, is significantly more peaked than the current profile before the disruption. In fact, it is found that the central current density can increase in spite of a reduction in the total current. (authors)

  5. Activation analysis of the compact ignition tokamak

    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.

  6. Activation analysis of the compact ignition tokamak

    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

  7. Effect of impurity radiation on tokamak equilibrium

    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)

  8. Mass spectrometry instrumentation in TN (Novillo Tokamak)

    The mass spectrophotometry in the residual gases analysis in high vacuum systems, in particular in the Novillo Tokamak (TN), where pressures are required to be of the order 10-7 Torr, is carried out through an instrumental support with infrastructure configured in parallel to the experimental planning in this device. In the Novillo as well as other Tokamaks, it is necessary to condition the vacuum chamber for improving the main discharge parameters. At the present time, in this Tokamak the conditioning quality is presented determined by means of a mass spectrophotometer. A general instrumental description is presented associated with the Novillo conditioning, as well as the spectras obtained before and after operation. (Author)

  9. Current drive by spheromak injection into a tokamak

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

    1990-01-01

    We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma (n¯3 increases by a factor of 6) then becomes hollow, suggestive of...

  10. Periodic disruptions in the MT-1 tokamak

    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

  11. Can better modelling improve tokamak control?

    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

  12. Tokamak power systems studies, FY 1985

    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.

  13. A method for tokamak neutronics calculations

    This paper presents a new method for neutron transport calculation in tokamak fusion reactors. The computational procedure is based on the solution of the even-parity transport equation in a toroidal geometry. The angular neutron distribution is treated by even-parity spherical harmonic expansion, while the spatial dependence is approximated by using R-function finite elements that are defined for regions of arbitrary geometric shape. In order to test the method, calculation of a simplified tokamak model is carried out. The results are compared with the results from the literature and for the same order of accuracy a reduction of the number of spatial unknowns is shown. (author)

  14. Electronic system of TBR tokamak device

    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)

  15. Tokamak Engineering Technology Facility scoping study

    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.

  16. Radial electric fields for improved tokamak performance

    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

  17. Tokamak Spectroscopy for X-Ray Astronomy

    Fournier, Kevin B.; Finkenthal, M.; Pacella, D.; May, M. J.; Soukhanovskii, V.; Mattioli, M.; Leigheb, M.; Rice, J. E.

    2000-01-01

    This paper presents the measured x-ray and Extreme Ultraviolet (XUV) spectra of three astrophysically abundant elements (Fe, Ca and Ne) from three different tokamak plasmas. In every case, each spectrum touches on an issue of atomic physics that is important for simulation codes to be used in the analysis of high spectral resolution data from current and future x-ray telescopes. The utility of the tokamak as a laboratory test bed for astrophysical data is demonstrated. Simple models generated with the HULLAC suite of codes demonstrate how the atomic physics issues studied can affect the interpretation of astrophysical data.

  18. Multichannel submillimeter interferometer for tokamak density measurements

    A two-channel, submillimeter (SMM) laser, electron-density interferometer has been operated successfully on the ISX tokamak. The interferometer is the first phase of a diagnostic system to measure the tokamak plasma current density using the Faraday rotation of the polarization vector of SMM laser beams. Deuterated formic acid lasers (lambda = 0.381 mm) have produced cw power of 10 mW. The interferometer has performed successfully for line-averaged electron densities as high as 8 x 1013 cm-3

  19. Tokamak power systems studies, FY 1985

    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

  20. Dynamics and Feedback Control of Plasma Equilibrium Position in a Tokamak.

    Burenko, Oleg

    A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems. The major parameters governing the plasma equilibrium position stability of a tokamak are shown to be (1) external magnetic field decay index, (2) transformer iron core effect, (3) plasma current, (4) radial rate-of-change inductance parameter, (5) vertical rate-of-change inductance parameter, and (6) vacuum vessel eddy-current time constant. An important and unique result is derived, showing that for a vacuum vessel eddy-current time constant exceeding a certain value the vertical plasma equilibrium position is stable, in spite of an intentional vertical instability design represented by a negative decay index. It is shown that a tokamak design having a theoretical set of positive decay index, negative radical rate-of-change inductance parameter, and positive vertical rate-of-change inductance parameter is expected to have a better plasma equilibrium position stability tolerance than a tokamak design having the same set with the signs reversed. The results of an actual hardware ISX-A tokamak plasma displacement feed-back control system design are presented. It is shown that a theoretical design computer

  1. DIII-D research operations

    The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R ampersand D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma

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

    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

  3. A Web-Based System for Remote Data Browsing in HT-7 Tokamak

    Cheng Ting; Luo Jiarong; Meng Yuedong; Wang Huazhong

    2005-01-01

    HT-7 is the first superconducting tokamak device for fusion research in China. Many experiments have been performed on the HT-7 tokamak since 1994 with numerous satisfactory results achieved in the fusion research field. As more and better communication is required with other fusion research laboratories, remote access to experimental data is becoming increasingly important in order to raise the degree of openness of experiments and to expand research results.The web-based remote data browsing system enables authorized users in geographically different locations to view and search for experimental data without having to install any utility software at their terminals. The three-tier software architecture and thin client technology are used to operate the system effectively. This paper describes the structure of the system and the realization of its functions, focusing on three main points: the communication between the participating tiers, the data structure of the system and the visualization of the raw data on web pages.

  4. Atomic data for integrated tokamak modelling

    . 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)

  5. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    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

  6. Microinstabilities in weak density gradient tokamak systems

    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

  7. High βp bootstrap tokamak reactor

    Basic characteristics of a steady state tokamak fusion reactor is presented. The minimum required energy multiplication factor Q is found to be 20 to 30 for the feasibility of the fusion reactor. Such a high Q steady state tokamak operation is possible, within our present knowledge of the operational constraints and the current drive physics, when a large fraction of the plasma current is carried by the bootstrap current. Operation at high βp (≥2.0) and high qψ (=4-5) with relatively small εβp (3) and fusion output power (2.5 GW) and is consistent with the present knowledges of the plasma physics of the tokamak, namely the Troyon limit, the energy confinement scalings, the bootstrap current, the current drive efficiency (NB current drive with the total power of 70 MW and the beam energy of 1 MeV) with a favorable aspect on the formation of the cold and dense diverter plasma-condition. From the economical aspect of the tokamak fusion reactor, a more compact reactor is favorable. The use of the high field magnet with Bmax = 16T (for example Ti-doped Nb3Sn conductor) enables to reduce the total machine size to 50% of the above-described conventional design, namely Rp = 7m, Vp = 760m-3, PF = 2.8 GW. (author)

  8. Tokamak fusion test reactor. Final design report

    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

  9. Advanced tokamak concepts and reactor designs

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  10. Plasma-gun fueling for tokamak reactors

    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

  11. Toroidal Alfven wave stability in ignited tokamaks

    Cheng, C.Z.; Fu, G.Y.; Van Dam, J.W.

    1989-01-01

    The effects of fusion-product alpha particles on the stability of global-type shear Alfven waves in an ignited tokamak plasma are investigated in toroidal geometry. Finite toroidicity can lead to stabilization of the global Alfven eigenmodes, but it induces a new global shear Alfven eigenmodes, which is strongly destabilized via transit resonance with alpha particles. 8 refs., 2 figs.

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

    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

  13. Analysis of sawtooth relaxation oscillations in tokamaks

    Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated

  14. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    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

    2003-10-01

    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.

  15. Slow bank system of SINP-Tokamak: A short report

    SINP Tokamak was made operational in July, 1987. The power supply system of the tokamak at that time was designed for a plasma duration of around 2 ms for a peak plasma current of 75 kA. Efforts were directed to increase this duration to 20 ms with the help of a slow bank system designed to work in conjunction with the original fast bank system. The design aspects of the system were completed and the system has been partially executed. Subsequent to this partial implementation, efforts were directed to incorporate the necessary control system and interface facilities between the existing fast bank and the developed slow bank systems. The significant features of the control circuits are that they work according to a well thought out sequences of logic and are designed to guard against possible failures in the existing or the developed power supplies. Efforts have been put to make the operation of the system as much user-friendly as could be worked out within certain practical constraints. The control circuit and interface facilities have been put to extensive tests and are found to work satisfactorily. The entire power supply system is now in active use for different research programmes in the group. (author)

  16. Observation of MHD phenomenon for SST-1 superconducting tokamak

    Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (major radius = 1.1 m, minor radius = 0.2 m) and is operational at the Institute for Plasma Research (IPR), India. In the last few experimental campaigns SST-1 has successfully achieved plasma current in order of 60-70 kA and plasma duration in excess of ∼ 500ms at a central magnetic field of 1.5T. An attempt has made to study the behavior of the magneto-hydrodynamic (MHD) activity during different phases of plasma pulse which leads to major/minor disruptions, its present modes (poloidal/toroidal mode number i.e. m=2, n=1) impact on plasma confinement and signature of lock mode and its frequency in the SST-1 plasma using experimental data from Mirnov signals. Observed MHD phenomenon has also been correlated with other diagnostics (i.e. ECE, Density, X-Ray etc.) and heating system (ECRH) for the recent campaigns of SST-1. (author)

  17. Measurement of Sheared Flows in the Edge Plasma of the CASTOR Tokamak

    Brotánková, Jana; Stöckel, Jan; Horáček, Jan; Seidl, Jakub; Ďuran, Ivan; Hron, Martin; Van Oost, G.

    2009-01-01

    Roč. 35, č. 11 (2009), s. 980-986. ISSN 1063-780X. [IAEA Technical Meeting on Research Using Small Fusion Devices/18th./. Alushta (Krym), 25.09.2008-27.09.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * probe diagnostics * shear ed flows * edge plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.584, year: 2009 http://www.springerlink.com/content/u571504gmq118314/

  18. Equilibrium system analysis in a tokamak ignition experiment

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies

  19. Equilibrium system analysis in a tokamak ignition experiment

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  20. Equilibrium system analysis in a tokamak ignition experiment. Final report

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  1. Maryland controlled fusion research program

    In this paper, we summarize the technical progress in four major areas of tokamak research: (a) L/H transition and edge turbulence and transport; (b) active control of microturbulence and transport; (c) major disruptions; and (d) the sawtooth crash

  2. Characterization of the Tokamak Novillo in cleaning regime; Caracterizacion del Tokamak Novillo en regimen de limpieza

    Lopez C, R.; Melendez L, L.; Valencia A, R.; Chavez A, E.; Colunga S, S.; Gaytan G, E

    1992-02-15

    In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip{sub t} like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I{sub (p)}t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)

  3. First experiments on the TO-2 tokamak with a divertor

    Long stable discharges have been obtained in a recetrack tokamak with toroidal divertors in low plasma density regime. Divertors sharply limit plasma filament cross section, plasma density decreasing by an order at 1 cm length near the separatrix. 8 mm thick well formed flux of plasma appears at the divertor plate. Divertor power efficiency at different modes of operation is 50- 70 %. As compared to the TO-1 nondivertor tokamak some plasma filament hot zone expansion is recorded in the TO-2 tokamak

  4. Banana orbits in elliptic tokamaks with hole currents

    Martin, P.; Castro, E.; Puerta, J.

    2015-03-01

    Ware Pinch is a consequence of breaking of up-down symmetry due to the inductive electric field. This symmetry breaking happens, though up-down symmetry for magnetic surface is assumed. In previous work Ware Pinch and banana orbits were studied for tokamak magnetic surface with ellipticity and triangularity, but up-down symmetry. Hole currents appear in large tokamaks and their influence in Ware Pinch and banana orbits are now considered here for tokamaks magnetic surfaces with ellipticity and triangularity.

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

    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

  6. Systems studies of high-field tokamak ignition experiments

    A study of the interaction between the physics of ignition and the engineering constraints in the design of compact, high-field tokamak ignition demonstration devices is presented. The studies investigate the effects the various electron and ion thermal diffusivities, which result from the many tokamak scaling laws, have on the design parameters of an ignition device and show the feasibility of building and igniting a compact tokamak (R<1m). The relevant machine technology is discussed

  7. Disruption generated secondary runaway electrons in present day tokamaks

    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

  8. OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM

    BURRELL,KH

    2002-11-01

    OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, the authors have made significant progress in developing the building blocks needed for AT operation: (1) the authors have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {le} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. They have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiation power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet

  9. Importance of the fine structure in a tokamak for the abnormal transport and the internal disruptions; Importance de la structure magnetique fine dans un Tokamak pour le transport anormal et les disruptions internes

    Sabot, R.

    1996-02-28

    The problem of energy transport in a Tokamak, in presence of magnetic islets, has been treated by decomposing this problem in different bricks. To assembly the different bricks the model of dynamic percolation, which couples by the intermediate of scattering coefficient, the activity of transport sites (islets size) to the profile of transported quantity (temperature profile) has been chosen. The results, got with this model, results connected to the hypothesis of a limited number of islets, agree with the different observations. A possible application of this model could be the exploration of different operating conditions of Tokamak and a research of improved confinement running. (N.C.). 149 refs., 85 figs.

  10. Numerical studies of edge localized instabilities in tokamaks

    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

  11. Design and construction of the KSTAR tokamak

    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)

  12. The tokamak - an imperfect frame of refernce

    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.)

  13. Magnetic sensor for steady state tokamak

    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).

  14. The physics of tokamak start-up

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

  15. Microinstability theory in tokamaks: a review

    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

  16. Rapidly Moving Divertor Plates In A Tokamak

    S. Zweben

    2011-05-16

    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.

  17. Runaway acceleration during magnetic reconnection in tokamaks

    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

  18. Rapidly Moving Divertor Plates In A Tokamak

    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.

  19. The Spherical Tokamak MEDUSA for Costa Rica

    Ribeiro, Celso; Vargas, Ivan; Guadamuz, Saul; Mora, Jaime; Ansejo, Jose; Zamora, Esteban; Herrera, Julio; Chaves, Esteban; Romero, Carlos

    2012-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R>=3.6, under design[2]) and also the ongoing activities in low temperature plasmas. Courses in plasma physics at undergraduate and post-graduate joint programme levels are regularly conducted. The scientific programme is intend to clarify several issues in relevant physics for conventional and mainly STs, including transport, heating and current drive via Alfv'en wave, and natural divertor STs with ergodic magnetic limiter[3,4]. [1] G.D.Garstka, PhD thesis, University of Wisconsin at Madison, 1997 [2] L.Barillas et al., Proc. 19^th Int. Conf. Nucl. Eng., Japan, 2011 [3] C.Ribeiro et al., IEEJ Trans. Electrical and Electronic Eng., 2012(accepted) [4] C.Ribeiro et al., Proc. 39^th EPS Conf. Contr. Fusion and Plasma Phys., Sweden, 2012

  20. Global migration of impurities in tokamaks

    The migration of impurities in tokamaks has been studied with the help of tracer-injection (13C and 15N) experiments in JET and ASDEX Upgrade since 2001. We have identified a common pattern for the migrating particles: scrape-off layer flows drive impurities from the low-field side towards the high-field side of the vessel. Migration is also sensitive to the density and magnetic configuration of the plasma, and strong local variations in the resulting deposition patterns require 3D treatment of the migration process. Moreover, re-erosion of the deposited particles has to be taken into account to properly describe the migration process during steady-state operation of the tokamak. (paper)

  1. KTM Tokamak operation scenarios software infrastructure

    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.

  2. KTM Tokamak operation scenarios software infrastructure

    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.

  3. Boundary Plasma Turbulence Simulations for Tokamaks

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

    2008-05-15

    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{sub e}; T{sub 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.

  4. Upgrade of plasma density feedback control system in HT-7 tokamak

    ZHAO Da-Zheng; LUO Jia-Rong; LI Gang; JI Zhen-Shan; WANG Feng

    2004-01-01

    The HT-7 is a superconducting tokamak in China used to make researches on the controlled nuclear fusion as a national project for the fusion research. The plasma density feedback control subsystem is the one of the subsystems of the distributed control system in HT-7 tokamak (HT7DCS). The main function of the subsystem is to control the plasma density on real-time. For this reason, the real-time capability and good stability are the most significant factors, which will influence the control results. Since the former plasma density feedback control system (FPDFCS) based on Windows operation system could not fulfill such requirements well, a new subsystem has to be developed. The paper describes the upgrade of the plasma density feedback control system (UPDFCS), based on the dual operation system (Windows and Linux), in detail.

  5. Electromagnetic effects of plasma disruptions in tokamaks

    The tokamak is modeled as typically 100 mutually-coupled toroidal circuits. The self and mutual inductances and the currents and voltages are calculated. Using the calculated currents, the poloidal magnetic field and the electromagnetic forces as functions of space and time are calculated. The major conclusion of the analysis is that the torus sectors should be electrically connected to each other near the plasma. Such connections reduce the structural loads, eliminate arcing, and reduce the induced potentials in the poloidal field coils

  6. Confinement scaling and ignition in tokamaks

    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) . 1015 cm-3, high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition

  7. Smaller coil systems for tokamak reactors

    Ripple reduction by ferro-magnetic iron shielding is used to reduce the size of the toroidal field coils down to 7.8 by 10.4 m bore for a commercial tokamak reactor design with plasma parameters similar to STARFIRE. For maximum effectiveness, it is found that the blocks of ferromagnetic iron shielding should have triangular cross section and should be placed as close to the plasma as possible

  8. Comparison of tokamak burn cycle options

    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

  9. Tore Supra. Basic design Tokamak system

    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

  10. Self-Organized Stationary States of Tokamaks.

    Jardin, S C; Ferraro, N; Krebs, I

    2015-11-20

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping." PMID:26636854

  11. Tokamak with liquid metal toroidal field coil

    Ohkawa, Tihiro; Schaffer, Michael 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.

  12. Lower hybrid heating of tokamaks to ignition

    The incorporation of a quasi-linear collisional wave damping model of lower hybrid electron heating into a radial transport code reveals favourable prospects for heating tokamak plasmas to ignition. The RF frequencies considered here are such that the wave interaction is primarily with the electrons. For a particular test reactor design, 30 MW of lower hybrid power used in conjunction with a programmed plasma density start-up suffices to initiate a self-sustained thermonuclear burn. (author)

  13. Fusion technology applications of the spherical tokamak

    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)

  14. Development of Atomic Beam Probe for tokamaks

    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#

  15. Self-Organized Stationary States of Tokamaks

    Jardin, S. C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ferraro, N. [General Atomics, San Diego, CA (United States); Krebs, I. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Max-Plank-Institut fur Plasmaphysik, Garching, Germany

    2015-11-01

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  16. MHD stability of an almost circular tokamak

    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

  17. Conditioning of the vacuum chamber of the Tokamak Novillo

    The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10-7 Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)

  18. Observation and Prediction of Runaway Electrons in the COMPASS Tokamak

    Papřok, R.; Krlín, Ladislav; Stöckel, Jan

    Vol. 2. Prague : MATFYZPRESS, 2013 - (Šafránková, J.; Pavlů, J.), s. 60-66 ISBN 978-80-7378-251-1. - (WDS). [Annual Conference of Doctoral Students – WDS 2013 /22./. Praha (CZ), 04.06.2013-07.06.2013] R&D Projects: GA MŠk 7G10072; GA MŠk LA08048 Grant ostatní: EURATOM(XE) FU07-CT-2007-00060 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * COMPASS Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/proc/pdf13/WDS13_209_f2_Paprok.pdf

  19. Neutral beam injection system design for KSTAR tokamak

    Choi, B.H.; Lee, K.W.; Chung, K.S.; Oh, B.H.; Cho, Y.S.; Bae, Y.D.; Han, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    The NBI system for KSTAR (Korean Superconducting Tokamak Advanced Research) has been designed based on conventional positive ion beam technology. One beam line consists of three ion sources, three neutralizers, one bending magnet, and one drift tube. This system will deliver 8 MW deuterium beam to KSTAR plasma in normal operation to support the advanced experiments on heating, current drive and profile control. The key technical issues in this design were high power ion source(120 kV, 65 A), long pulse operation (300 seconds; world record is 30 sec), and beam rotation from vertical to horizontal direction. The suggested important R and D points on ion source and beam line components are also included. (author). 7 refs., 27 figs., 1 tab.

  20. Vlasov tokamak equilibria with shearad toroidal flow and anisotropic pressure

    Throumoulopoulos, George; Kuiroukidis, Apostolos; Tasso, Henri

    2015-11-01

    By choosing appropriate deformed Maxwellian ion and electron distribution functions depending on the two particle constants of motion, i.e. the energy and toroidal angular momentum, we reduce the Vlasov axisymmetric equilibrium problem for quasineutral plasmas to a transcendental Grad-Shafranov-like equation. This equation is then solved numerically under the Dirichlet boundary condition for an analytically prescribed boundary possessing a lower X-point to construct tokamak equilibria with toroidal sheared ion flow and anisotropic pressure. Depending on the deformation of the distribution functions these steady states can have toroidal current densities either peaked on the magnetic axis or hollow. These two kinds of equilibria may be regarded as a bifurcation in connection with symmetry properties of the distribution functions on the magnetic axis. This work has received funding from (a) the National Programme for the Controlled Thermonuclear Fusion, Hellenic Republic, (b) Euratom research and training programme 2014-2018 under grant agreement No 633053.

  1. Preliminary design study of a steady state tokamak device

    Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

  2. Management and protection system for superconducting tokamak

    Juszczyk, B.; Wojenski, A.; Zienkiewicz, P.; Kasprowicz, G.; Pozniak, K.; Romaniuk, R.

    2015-09-01

    This paper describes system for a diagnostics of a high-voltage power supply section of tokamaks. System is designed to assure reliability and safety of power supply subsystems. It is divided into two main components: remote and local. Remote part is located near tokamak, whereas local part can be localised away from the tokamak area. The remote side consists of custom, standalone devices. On the other hand, the local device is based on the uTCA.4 architecture. Components are connected with an optic fibre over a link-layer protocol which provides high throughput, low latency and transmission redundancy. All main operations ie. data processing, transmission etc. are performed on the FPGA devices. At the local side there is one device treated as a master device. It implements sort of a routing table which connects consecutive system inputs and outputs. It also provides possibility for some user defined data processing. This document contains general system overview, short description of hardware used in the project and gateware implementation.

  3. Experimental and theoretical basis for advanced tokamaks

    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

  4. The minimum dissipation state for tokamaks

    The principle of minimum dissipation rate subject to helicity and energy balance is applied to tokamaks with an arbitrary aspect ratio. We solved the resulting Euler-Lagrange equations analytically and numerically. It is found that for low and general aspect ratio tokamaks, there exists different typical minimum dissipation state, corresponding to the typical experimental current profile respectively. It is also found that there exist different types of relaxed states in different regions of the parameter space for a selected device. Three forms of current profile are presented under different experimental conditions for a low aspect ratio tokamak like NSTX. The first peaks in the edge region of the high field side similar as the typical experimental form. The second peaks in the central region on the equatorial plane. The third may have a hole or reverse in the central part. E0/ηB0 is the key parameter in determining the final relaxed state; both the second and the third states could be obtained violently by increasing it to be above a critical value. (author)

  5. Electron cyclotron emission from tokamak plasmas

    Emitted electron radiation can be used as a diagnostic signal to measure the electron temperature of a thermonuclear plasma. This type of diagnostics is well established in tokamak physics. In ch. 2 of this thesis the development, calibration and special design features are treated of a six-channel prototype of a twelve-channel grating spectrometer which is built for JET at Culham for electron cyclotron emission (ECE) measurements. In order to test this prototype measurements have been performed with the T-10 tokamak at the Kurchatov Institute in Moscow. With this prototype nearly half of the temperature profile of the T-10 could be measured. Detailed observations of sawteeth instabilities have been performed. Plasma heating by electron cyclotron resonance heating experiments was studied. A detailed description of these measurements and results is given in ch. 3. Often ECE spectra from tokamaks showed non-thermal features. In order to interprete them a computer code Notec has been developed. This code that calculates the ECE radiation emerging from the plasma for a 3-D configuration, is described in ch. 4. Some preliminary results and applications are presented. (Auth.)

  6. The Spherical Tokamak MEDUSA for Mexico

    Ribeiro, C.; Salvador, M.; Gonzalez, J.; Munoz, O.; Tapia, A.; Arredondo, V.; Chavez, R.; Nieto, A.; Gonzalez, J.; Garza, A.; Estrada, I.; Jasso, E.; Acosta, C.; Briones, C.; Cavazos, G.; Martinez, J.; Morones, J.; Almaguer, J.; Fonck, R.

    2011-10-01

    The former spherical tokamak MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R Mexico, as part of an agreement between the Faculties of Mech.-Elect. Eng. and Phy. Sci.-Maths. The main objective for having MEDUSA is to train students in plasma physics & technical related issues, aiming a full design of a medium size device (e.g. Tokamak-T). Details of technical modifications and a preliminary scientific programme will be presented. MEDUSA-MX will also benefit any developments in the existing Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

  7. Nonlinear simulation studies of tokamaks and STs

    The multilevel physics, massively parallel plasma simulation code, M3D has been used to study spherical toris (STs) and tokamaks. The magnitude of outboard shift of density profiles relative to electron temperature profiles seen in NSTX under strong toroidal flow is explained. Internal reconnection events in ST discharges can be classified depending on the crash mechanism, just as in tokamak discharges; a sawtooth crash, disruption due to stochasticity, or high-β disruption. Toroidal shear flow can reduce linear growth of internal kink. It has a strong stabilizing effect nonlinearly and causes mode saturation if its profile is maintained, e.g. through a fast momentum source. Normally, however, the flow profile itself flattens during the reconnection process, allowing a complete reconnection to occur. In some cases, the maximum density and pressure spontaneously occur inside the island and cause mode saturation. Gyrokinetic hot particle/MHD hybrid studies of NSTX show the effects of fluid compression on a fast-ion driven n = 1 mode. MHD studies of recent tokamak experiments with a central current hole indicate that the current clamping is due to sawtooth-like crashes, but with n = 0. (author)

  8. Nonlinear simulation studies of tokamaks and ST's

    The multilevel physics, massively parallel plasma simulation code, M3D has been used to study ST's and tokamaks. The magnitude of outboard shift of density profiles relative to electron temperature profiles seen in NSTX under strong toroidal flow is explained. IRE's in ST discharges can be classified depending on the crash mechanism, just as in tokamak discharges; a sawtooth crash, disruption due to stochasticity, or high-β disruption. Toroidal shear flow can reduce linear growth of internal kink. It has a strong stabilizing effect nonlinearly and causes mode saturation if its profile is maintained, e.g., through a fast momentum source. Normally however, the flow profile itself flattens during the reconnection process, allowing a complete reconnection to occur. In some cases, the maximum density and pressure spontaneously occur inside the island and cause mode saturation. Gyrokinetic hot particle/MHD hybrid studies of NSTX show the effects of fluid compression on a fast-ion-driven n=1 mode. MHD studies of recent tokamak experiments with a central current hole indicate that the current clamping is due to sawtooth-like crashes, but with n=0. (author)

  9. Collisionless microtearing modes in standard tokamak configurations

    Microtearing Modes (MTM) are electromagnetic microinstabilities occurring in magnetically confined fusion plasmas driven by parallel electron current and collisions in the presence of electron temperature gradient. MTMs were first predicted to occur in such plasmas in early 70s. Collisional MTMs have recently gathered attention in Spherical Tokamak configurations and RFPs. Very recently collisional MTMs have been reported in configuration relevant to standard tokamak, namely ASDEX-U. Perhaps for the first time, we show the existence of MTMs in purely collisionless limit and in large aspect ratio tokamak configurations using fully gyrokinetic full radius linear calculations. The physics of both electron scale as well as minor radius scale are resolved in the studies. Results of the studies, such as the 2-D structure of the mode and the dependence of growth rates on plasma pressure, perpendicular (to B0) wavelength spectrum and the effect of Landau damping and magnetic drift resonance will be presented. A comparison with another electromagnetic mode, namely Kinetic Ballooning Mode, which is driven by ion temperature gradient will also be shown. (author)

  10. Design of the ITER tokamak assembly tools

    ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40 deg. sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values

  11. ECH on the MTX [Microwave Tokamak Experiment

    The Microwave Tokamak Experiment (MTX) at LLNL is investigating the heating of high density Tokamak plasmas using an intense pulse FEL. Our first experiments, now beginning, will study the absorption and plasma heating of single FEL pulses (20 ns pulse length and peak power up to 2 GW) at a frequency of 140 GHz. A later phase of experiments also at 140 GHz will study FEL heating at 5 kHz rate for a pulse train up to 50 pulses (35 ns pulse length and peak power up to 4 GW). Future operations are planned at 250 GHz with an average power of 2 MW for a pulse train of 0.5 s. The microwave output of the FEL is transported quasi-optically to the tokamak through a window-less, evacuated pipe of 20 in. diameter, using a six mirror system. Computational modelling of the non-linear absorption for the MTX geometry predicts single-pass absorption of 40% at a density and temperature of 1.8 /times/ 1020m/sup /minus/3/ and 1 keV, respectively. To measure plasma microwave absorption and backscatter, diagnostics are available to measure forward and reflected power (parallel wire grid beam-splitter and mirror directional couplers) and power transmitted through the plasma (segmented calorimeter and waveguide detector). Other fast diagnostics include ECE, Thompson scattering, soft x-rays, and fast magnetic probes. 8 refs., 2 figs

  12. Recent results from the DIII-D tokamak

    Petersen, P.I.

    1998-02-01

    The DIII-D national fusion research program focuses on establishing the scientific basis for optimization of the tokamak approach to fusion energy production. The symbiotic development of research, theory, and hardware continues to fuel the success of the DIII-D program. During the last year, a radiative divertor and a second cryopump were installed in the DIII-D vacuum vessel, an array of central and boundary diagnostics were added, and more sophisticated computer models were developed. These new tools have led to substantial progress in the understanding of the plasma. The authors now have a better understanding of the divertor as a means to manage the heat, particle, and impurity transport pumping of the plasma edge using the in situ divertor cryopumps effectively controls the plasma density. The evolution of diagnostics that probe the interior of the plasma, particularly the motional Stark effect diagnostic, has led to a better understanding of the core of the plasma. This understanding, together with tools to control the profiles, including electron cyclotron waves, pellet injection, and neutral beam injection, has allowed them to progress in making plasma configurations that give rise to both low energy transport and improved stability. Most significant here is the use of transport barriers to improve ion confinement to neoclassical values. Commissioning of the first high power (890 kW) 110 GHz gyrotron validates an important tool for managing the plasma current profile, key to maintaining the transport barriers. An upgraded plasma control system, ``isoflux control,`` which exploits real time MHD equilibrium calculations to determine magnetic flux at specified locations within the tokamak vessel and provides the means for precisely controlling the plasma shape and, in conjunction with other heating and fueling systems, internal profiles.

  13. The role of radial electric fields in tokamaks TEXTOR-94, CASTOR, and T-10

    Van Oost, G.; Gunn, J. P.; Melnikov, A.; Stöckel, Jan; Tendler, M.

    2001-01-01

    Roč. 51, č. 10 (2001), s. 957-975. ISSN 0011-4626. [Europhysics Workshop on The Role Electric Fields in Plasma Confinement and Exhaust/4th./. Funchal, Madeira, 24.06.2001-25.06.2001] Institutional research plan: CEZ:AV0Z2043910 Keywords : electric fields, tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.345, year: 2001

  14. Validity of Self-Organized Criticality model for the CASTOR tokamak edge plasmas

    Ďuran, Ivan; Stöckel, Jan; Horáček, Jan; Jakubka, Karel; Kryška, Ladislav; Hron, Martin

    volume 24B. Mulhouse : European Physical Society, 2000 - (Szegö, K.; Todd, T.; Zoletnik, S.), s. 1693-1696 - (Europhysics Conference Abstracts.. 24B). [European Physical Society Conference on Controlled Fusion and Plasma Physics/27th./. Budapest (HU), 12.06.2000-16.06.2000] Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, plasma Subject RIV: BL - Plasma and Gas Discharge Physics

  15. The optimization of resonant magnetic perturbation spectra for the COMPASS tokamak

    Cahyna, Pavel; Pánek, Radomír; Fuchs, Vladimír; Krlín, Ladislav; Bécoulet, M.; Nardon, E.; Huysmans, G.

    2009-01-01

    Roč. 49, č. 5 (2009), 055024 1-055024 7. ISSN 0029-5515. [IAEA Fusion Energy Conference/22nd./. Geneva, 13.10.2008-18.10.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : resonant magnetic perturbations * ELM control * magnetic islands * saddle coils * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://dx.doi.org/10.1088/0029-5515/49/5/055024

  16. Low cost alternative of high speed visible light camera for tokamak experiments

    Odstrčil, T.; Odstrčil, Michal; Grover, O.; Svoboda, V.; Ďuran, Ivan; Mlynář, Jan

    2012-01-01

    Roč. 83, č. 10 (2012), 10E505-10E505. ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostic * high speed camera * GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://dx.doi.org/10.1063/1.4731003

  17. Quasi-neutral simulations of tokamak scrape-off layer currents

    Fuchs, Vladimír; Gunn, J. P.

    Varšava, 2007. P-4.033. ISBN 2-914771-40-1. [European Physical Society Conference on Plasma Physics/34th./. 2.7.2007-6.7.2007, Varšava] R&D Projects: GA ČR(CZ) GA202/04/0360 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak edge plasma * PIC simulations Subject RIV: BL - Plasma and Gas Discharge Physics http://www.eps.ifpilm.waw.pl/abstracts_all.zip

  18. Landau damping of the lower hybrid grill spectrum by tokamak edge electrons

    Fuchs, Vladimír; Gunn, J. P.; Petržílka, Václav; Horáček, Jan; Seidl, Jakub; Ekedahl, A.; Goniche, M.; Hillairet, J.

    Vol. 1187. Melville : American Institute of Physics, 2009 - (Bobkov, V.; Noterdaeme, J.), s. 383-386 ISBN 978-0-7354-0753-4. - (AIP Conference Proceedings. 1187). [Topical Conference on Radio Frequency Power in Plasmas/18th./. Ghent (BE), 24.06.2009-26.06.2009] R&D Projects: GA ČR GA202/07/0044 Institutional research plan: CEZ:AV0Z20430508 Keywords : Lower hybrid wave * tokamak * plasma Subject RIV: BL - Plasma and Gas Discharge Physics

  19. Magnetic measurements using array of integrated Hall sensors on the CASTOR tokamak

    Ďuran, Ivan; Hronová-Bilyková, Olena; Stöckel, Jan; Sentkerestiová, J.; Havlíček, Josef

    2008-01-01

    Roč. 79, č. 10 (2008), 10F123-10F123. ISSN 0034-6748. [Topical Conference on High-Temperature Plasma Diagnostics/17th./. Albuquerque, 11.05.2008-15.05.2008] R&D Projects: GA MPO 2A-1TP1/101 Institutional research plan: CEZ:AV0Z20430508 Keywords : Galvanomagnetic Sensor * Fusion Reactor * Magnetic Diagnostics * CASTOR tokamak Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.738, year: 2008

  20. Effects of orbit squeezing on neoclassical toroidal plasma viscosity in tokamaks

    Shaing, K.C.; Sabbagh, S.A.; Chu, M.S.; Bécoulet, M.; Cahyna, Pavel

    2008-01-01

    Roč. 15, č. 8 (2008), 082505-1-082505-8. ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma instability * plasma magnetohydrodynamics * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2965146

  1. Intermediate frequency band digitized high dynamic range radiometer system for plasma diagnostics and real-time Tokamak control

    Bongers, WA.; Van Beveren, V.; Thoen, D.J.; Nuij, P.J.W.M.; De Baar, M.R.; Donné, A.J.H.; Westerhof, E.; Goede, A.P.H.; Krijger, B.; Van den Berg, M.A.; Kantor, M.; Graswinckel, M.F.; Hennen, B.A.; Schüller, F.C.

    2011-01-01

    An intermediate frequency (IF) band digitizing radiometer system in the 100–200 GHz frequency range has been developed for Tokamak diagnostics and control, and other fields of research which require a high flexibility in frequency resolution combined with a large bandwidth and the retrieval of the f

  2. Conditioning of the vacuum chamber of the Tokamak Novillo; Acondicionamiento de la camara de vacio del Tokamak Novillo

    Valencia A, R.; Lopez C, R.; Melendez L, L.; Chavez A, E.; Colunga S, S.; Gaytan G, E

    1992-03-15

    The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10{sup -7} Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)

  3. Deposit of thin films for Tokamaks conditioning

    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 Zef 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

  4. Material erosion and migration in tokamaks

    Material migration is one of the outstanding issues facing successful long pulse, high power tokamak operation, both for the next step device, ITER, and the longer term economic and technological viability of fusion power. Erosion of tokamak first wall surfaces may occur via a number of processes, both steady state and transient, the relative importance of each of which depends sensitively on the nature of the driving mechanism and the wall material itself. The subsequent transport of this eroded material through the plasma and its redeposition, often in locations remote from the point of release, constitute the foundation of material migration. Such material movement is intimately linked with the critical issue of tritium retention (via the process of co-deposition), which, in ITER and beyond, will determine the duration over which the tokamak may be operated before removal of the retained fraction is imposed by nuclear safety restrictions. Of the three processes: erosion, large-scale material transport and co-deposition, transport is currently the least understood, leading to large uncertainties in the predicted T-retention in ITER, independently of the chosen wall materials. The low duty cycle and reduced energy and particle fluxes to first wall surfaces in today's tokamaks mean that the phenomena of migration is of no practical consequence to their operation. In steady state reactor-class devices, however, annual migration rates are currently predicted to be in the range of tons, even in the absence of transient events. These estimates are nevertheless associated with considerable uncertainty and, although the situation is unlikely to be completely resolved by the time ITER is constructed, a clearer understanding of the global migration picture is emerging from ongoing physics studies in current devices. In particular, the influential role of erosion at main chamber surfaces, followed by subsequent transfer to the divertor and the delicate erosion

  5. Turbulence, transport and confinement: from tokamaks to star magnetism

    This thesis is part of the general study of self-organization in hot and magnetized plasmas. We focus our work on two specific objects: stars and tokamaks. We use first principle numerical simulations to study turbulence, transport and confinement in these plasmas. The first part of this thesis introduces the main characteristics of stellar and tokamak plasmas. The reasons for studying them together are properly detailed. The second part is focused on stellar aspects. We study the interactions between the 3D turbulent motions in the solar convection zone with an internal magnetic field in the tachocline (the transition region between the instable and stable zones in the Sun). The tachocline is a very thin layer (less than five percent of the solar radius) that acts as a transport barrier of angular momentum. We show that such an internal magnetic field is not likely to explain the observed thickness of the tachocline and we give some insights on how to find alternative mechanisms to constrain it. We also explore the effect of the environment of star on its structure. We develop a methodology to study the influence of stellar wind and of the magnetic coupling of a star with its orbiting planets. We use the same methodology to analyse the magnetic interaction between a stellar wind and a planetary magnetosphere that acts as a transport barrier of matter. Then, the third part is dedicated to fusion oriented research. We present a numerical investigation on the experimental mechanisms that lead to the development of transport barriers in the plasma. These barriers are particularly important for the design of high performance fusion devices. The creation of transport barriers is obtained in turbulent first principle simulations for the very first time. The collaboration between the two scientific teams lead to the results presented in the fourth part of this thesis. An original spectral method is developed to analyse the saturation of stellar convective dynamos and of

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

    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. Fokker-Planck/Transport model for neutral beam driven tokamaks

    The application of nonlinear Fokker-Planck models to the study of beam-driven plasmas is briefly reviewed. This evolution of models has led to a Fokker-Planck/Transport (FPT) model for neutral-beam-driven Tokamaks, which is described in detail. The FPT code has been applied to the PLT, PDX, and TFTR Tokamaks, and some representative results are presented

  8. Tokamak plasma self-organization-synergetics of magnetic trap plasmas

    Razumova, K. A.; Andreev, V. F.; Eliseev, L. G.; Kislov, A. Y.; La Haye, R. J.; Lysenko, S. E.; Melnikov, A. V.; Notkin, G. E.; Pavlov, Y. D.; Kantor, M. Y.

    2011-01-01

    Analysis of a wide range of experimental results in plasma magnetic confinement investigations shows that in most cases, plasmas are self-organized. In the tokamak case, it is realized in the self-consistent pressure profile, which permits the tokamak plasma to be macroscopically MHD stable. Existin

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

    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

  10. Advanced tokamak operating modes in TPX and ITER

    A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER

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

    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.

  12. Lower hybrid heating experiments in tokamaks: an overview

    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

  13. Experimental data base of Tokamak KTM physical diagnostics

    The process of software creation of experimental data storage of Tokamak KTM physical diagnostics based on analysis of storage methods of operating Tokamaks data is considered. Task of specific kinds of information storage is solved; experimental data base that is thr part of system providing information analysis performance in the post-start period is developed.(author)

  14. Experimental studies of tokamak plasma in IPP Prague

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR during recent years. At present, investigation is primarily aimed at the anomalous transport and plasma-wall interaction in the tokamak under conditions of combined OH/LHCD regimes. Moreover, some New diagnostic methods were also developed and certain improvements in the CASTOR performance were achieved. (author). 41 refs

  15. Desirable engineering features of the next-generation tokamak device

    Recent scoping studies examined a series of superconducting, long-pulse Driven Current Tokamak (DCT) devices. One class of options is an ignited, D-T burning device designated DCT-8. It was concluded that the DCT-8 is a most attracttive engineering option to adequately bridge the gap between the Tokamak Fusion Test Reactor (TFTR) and the Engineering Test Reactor

  16. Tokamak Plasmas : Measurement of temperature fluctuations and anomalous transport in the SINP tokamak

    R Kumar; S K Saha

    2000-11-01

    Temperature fluctuations have been measured in the edge region of the SINP tokamak. We find that these fluctuations have a comparatively high level (30–40%) and a broad spectrum. The temperature fluctuations show a quite high coherence with density and potential fluctuations and contribute considerably to the anomalous particle flux.

  17. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    E. Perry; J. Chrzanowski; K. Rule; M. Viola; M. Williams; R. Strykowsky

    1999-11-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling.

  18. Experiences on vacuum conditioning in the cryostat of KSTAR tokamak

    Highlights: ► The vacuum of the cryostat has been stably maintained during the KSTAR operation. ► The detected cold leak at the PF/CS coils and CS structure. ► The present helium leak makes no issue for the cryostat operation. -- Abstract: Korea Superconducting Tokamak Advanced Research (KSTAR) device has been successfully operated for the plasma experiment from KSTAR 1st campaign to 4th campaign. The main pumping system for the cryostat has to maintain the target pressure below 1.0 × 10−4 mbar at room temperature and 1.0 × 10−5 mbar at extremely low temperature for the plasma experiment against the air leak coming from ports of vessel and/or the helium leak from cooling loops in the cryostat. No leak has been detected at room temperature. Unexpectedly, the cold-leak appeared in the cryostat at temperature around 50 K during the cool-down in the KSTAR 2nd campaign. We carefully analyzed the characteristics of detected cold leak because it can cause the increase of the base pressure in the cryostat. After the cool-down, the leak detection was performed to locate the position and size of the leak by the pressurizing the loops. As a result, it is found that the cold leak was located at cooling loops for PF/CS coils and CS structure. Nevertheless, the vacuum inside the cryostat was well maintained below 6.0 × 10−8 mbar during the entire operation period. The impact of the He-leak in present status on the plasma operation is negligible. However, we have found that the leak rate increases as a function of time. Therefore careful monitoring on cold-leak is an important technical issue for the operation of superconducting tokamak

  19. System assessment of helical reactors in comparison with tokamaks

    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-βN tokamak reactors. (author)

  20. High performance operational limits of tokamak and helical systems

    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)

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

    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

  2. Observation of ICRF [ion cyclotron range of frequencies] wave-packet propagation in a tokamak plasma

    Experimental observation of ICRF wave-packet propagation in a tokamak plasma is reported. Studies were carried out in the Caltech Research Tokamak in a pure hydrogen plasma and in a regime where fast-wave damping was sufficiently small to permit multiple toroidal transits of the wave-packet. Waves were launched by exciting a small loop antenna with a short burst of rf current and were detected with shielded magnetic probes. Probe scans revealed a large increase in wave-packet amplitude at smaller minor radii, and the packet velocity was found to be independent of radial position. Measurement of the packet transit time yielded direct information about the wave group velocity. Packet velocity was investigated as a function of the fundamental excitation frequency, plasma density, and toroidal magnetic field. Results are compared with the predictions of a cold plasma model which includes a vacuum layer at the edge. 24 refs., 8 figs

  3. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    Chektybayev, B., E-mail: Chektybaev@nnc.kz; Shapovalov, G.; Kolodeshnikov, A. [Institute of Atomic Energy Branch of National Nuclear Center, Kurchatov (Kazakhstan)

    2015-05-15

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  4. Fusion plasma theory. Task III. Auxiliary heating in tokamaks and tandem mirrors. Final report

    The research we have accomplished with this contract has focused on ICRF coupling, heating and breakeven studies for tokamaks and ECRF fundamental and second harmonic heating in tandem mirrors. The highlights include reviewed publication of ICRF Fokker-Planck heating and breakeven studies with international collaboration with the JET group, fundamental work on a differential equation for wave fields and a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas and a formulation and code development of slab matrix and differential equation solutions for ICRF waveguide coupling in tokamak edge regions. ECRF ray tracing studies have been carried out, and a reviewed paper published for fundamental and second harmonic propagation, absorption and whistler microinstabilities in tandem mirror plug and barrier regions of Phaedrus, TMX-U and TASKA

  5. The Tokamak Fusion Test Reactor D-T modifications and operations

    This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

  6. DAMAVAND - An Iranian tokamak with a highly elongated plasma cross-section

    The ''DAMAVAND'' facility is an Iranian Tokamak with a highly elongated plasma cross-section and with a poloidal divertor. This Tokamak has the advantage to allow the plasma physics research under the conditions similar to those of ITER magnetic configuration. For example, the opportunity to reproduce partially the plasma disruptions without sacrificing the studies of: equilibrium, stability and control over the elongated plasma cross-section; processes in the near-wall plasma; auxiliary heating systems, etc. The range of plasma parameters, the configuration of ''DAMAVAND'' magnetic coils and passive loops, and their location within the vacuum chamber allow the creation of the plasma at the center of the vacuum chamber and the production of two poloidal volumes (upper and lower) for the divertor. (author)

  7. Understanding and Control of Transport in Advanced Tokamak Regimes in DIII-D

    Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection takes place when the heating power exceeds a threshold value of about 9 MW, contrasting to CO-NBI injection, where Pthreshold NH89 = 9 for 16 confinement times has been accomplished in a discharge combining an ELMing H-mode edge and an ITB, and exhibiting ion thermal transport down to 2-3 times neoclassical. The microinstabilities usually associated with ion thermal transport are predicted stable, implying that another mechanism limits performance. High frequency MHD activity is identified as the probable cause

  8. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle.

    Chektybayev, B; Shapovalov, G; Kolodeshnikov, A

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors. PMID:26026523

  9. Atomic physics studies of highly charged ions on tokamaks using x-ray spectroscopy

    Beiersdorfer, P.; von Goeler, S.; Bitter, M.; Hill, K.W.

    1989-07-01

    An overview is given of atomic physics issues which have been studied on tokamaks with the help resolution x-ray spectroscopy. The issues include the testing of model calculations predicting the excitation of line radiation, the determination of rate coefficients, and accurate atomic structure measurements. Recent research has focussed primarily on highly charged heliumlike (22 less than or equal to Z less than or equal to 28) and neonlike (34 less than or equal to Z less than or equal to 63) ions, and results are presented from measurements on the PLT and TFTR tokamaks. Many of the measurements have been aided by improved instrumental design and new measuring techniques. Remarkable agreement has been found between measurements and theory in most cases. However, in this review those areas are stressed where agreement is worst and where further investigations are needed. 19 refs., 13 figs., 2 tabs.

  10. Atomic physics studies of highly charged ions on tokamaks using x-ray spectroscopy

    An overview is given of atomic physics issues which have been studied on tokamaks with the help resolution x-ray spectroscopy. The issues include the testing of model calculations predicting the excitation of line radiation, the determination of rate coefficients, and accurate atomic structure measurements. Recent research has focussed primarily on highly charged heliumlike (22 ≤ Z ≤ 28) and neonlike (34 ≤ Z ≤ 63) ions, and results are presented from measurements on the PLT and TFTR tokamaks. Many of the measurements have been aided by improved instrumental design and new measuring techniques. Remarkable agreement has been found between measurements and theory in most cases. However, in this review those areas are stressed where agreement is worst and where further investigations are needed. 19 refs., 13 figs., 2 tabs

  11. MHD stability of advanced tokamak scenarios

    Tokamak plasmas with a non-monotonic q-profile (current profile) and negative shear in the plasma centre have been associated with improved confinement and large pressure gradients in the region of negative shear. In JET, this regime, has been obtained with pellet injection (the PEP mode) and in DIII-D by ramping the plasma elongation. In JET, the phase of improved confinement is transient and usually ends in a collapse due to an MHD instability which leads to a redistribution of the current and a monotonic q-profile. The infernal mode, which is driven by a large pressure gradient in the region of low shear near the minimum in the q-profile, is the most likely candidate for the observed instability. To extend the transient phase to steady state, control of the shape of the current density profile is essential. The modelling of these advanced tokamak scenarios with a non-monotonic q-profile using non-inductive current drive of lower hybrid waves, fast waves, and neutral beams is discussed elsewhere. The aim is to find suitable initial states and to maintain MHD stability when the plasma β is built up. For this purpose, the robustness of the MHD stability of these configurations is studied with respect to changes in the position and in the depth of the minimum in q, and in the shape of the q and pressure profile. The classes of equilibria chosen for the analysis are based on the modelling of the current-drive schemes for advanced tokamak scenarios in JET. The toroidal ideal and resistive MHD stability code CASTOR is used for the stability calculations. (author) 7 refs., 4 figs

  12. Mathematical modeling plasma transport in tokamaks

    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%

  13. The physics of an ignited tokamak

    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

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

    The D III-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: 1) We have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; 2) Using this rotational stabilization, we have achieved βNH89 ≥ 10 for 4 τE limited by the neoclassical tearing mode; 3) Using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased βT by 60%; 4) We have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; 5) We have made the first integrated AT demonstration discharges with current profile control using ECCD; 6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and 7) We have demonstrated stationary tokamak operation for 6.5 s (36 τE) at the same fusion gain parameter of βNH89/q952 ≅ 0.4 as ITER but at much higher q95 = 4.2. We have developed general improvements applicable to conventional and advanced tokamak operating modes: 1) We have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 τE) with constant density and constant radiated power; 2) We have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; 3) We have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. (author)

  15. ECRH current drive in tokamak plasmas

    The current drive by electron cyclotron resonance heating (ECRH) is investigated in a typical magnetic field of tokamak with circular cross section. The trapped electrons and the modification of electron-cyclotron resonance condition by the relativistic mass increase are shown to have significant effects on the efficiency of this current drive. The efficiency strongly depends on the values of the parallel velocity u0 of resonant electrons, the inverse aspect ratio ε, the poloidal angle θ0 of absorption point, and the relativistic parameter S, which represents the strength of the relativistic correction to the resonance condition. (author)

  16. Scaling studies of beam-heated tokamaks

    Parametric scaling of neutral beam-heated tokamaks is examined to determine the trade-off between beam energy and power. It is shown that over a wide range of plasma parameters and assumed transport properties, the center mean plasma temperature is a function of P/sub A/E/sub B//sup delta/, where E/sub B/ and P/sub A/ are the beam energy and power per unit area, respectively, and delta is a calculable constant of order unity

  17. Bootstrap current estimate in the ETE Tokamak

    First estimates of the bootstrap current in the ETE small aspect ratio tokamak using the Hirshman single ion collisionless model show that we can expect from 25 to 55% of total bootstrap current depending on the optimization level of the plasma parameter profiles. Higher levels of bootstrap current are limited by peaked pressure profiles and βpol values which must be kept under a critical level due to stability conditions. Different methods for the trapped particle fraction calculation are also illustrated in this paper. (author). 7 refs., 5 figs., 1 tab

  18. Density measurement systems at SST Tokamak

    Electromagnetic wave experiences a phase difference while passing through the plasma with respect to the reference arm. This phase information gives line averaged electron plasma density. At SST-1 Tokamak, two microwave interferometer systems - (1) 100 GHz homodyne system and (2) 140 GHz phase locked heterodyne system, have been designed, developed and installed. In this paper developed systems performances as well as measurement descriptions are explained. A comparative study has been done to understand the measurement capabilities of the two independent systems and a good agreement is obtained. The measured density of the recent plasma discharges after first wall installation is in the range of 2 - 5 x 1012/ cm3. (author)

  19. Magnetic microtearing coherence in tokamak plasmas

    The analyses of the microtearing-modes coherence is effected. The tokamak characteristics, concerning fusion, electromagnetic confinement and turbulence are reviewed. The nature of the tearing modes, the variational principle of linear mode studies, a linear study in collisional and non-collisional plasma conditions are summarized, before studying the microtearing-mode coherence. The flux line configuration in the presence of a magnetic turbulence, the plasma response to a microtearing perturbation and instability, in the presence of a radial-electrons diffusion, is described. The autocoherence of microtearing modes in non-linear conditions are analyzed

  20. Infrared Thermography on the COMPASS Tokamak

    Vondráček, Petr; Horáček, Jan; Cahyna, Pavel; Pánek, Radomír; Uličný, J.

    Vol. 2. Prague : MATFYZPRESS, 2013 - (Šafránková, J.; Pavlů, J.), s. 80-85 ISBN 978-80-7378-251-1. - (WDS). [Annual Conference of Doctoral Students – WDS 2013 /22./. Praha (CZ), 04.06.2013-07.06.2013] R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Plasma diagnostics * infrared camera * tokamak Subject RIV: BL - Plasma and Gas Discharge Physic s http://www.mff.cuni.cz/veda/konference/wds/proc/pdf13/WDS13_212_f2_Vondracek.pdf