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Sample records for aditya tokamak

  1. 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

  2. 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)

  3. 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

  4. 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.

  5. 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.

  6. 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)

  7. 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)

  8. 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)

  9. 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)

  10. 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)

  11. 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.

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. 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)

  17. 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)

  18. 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)

  19. 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)

  20. 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.

  1. 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

  2. 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)

  3. 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

  4. 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.

  5. 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)

  6. 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)

  7. 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)

  8. 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)

  9. 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.

  10. 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)

  11. 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)

  12. 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).

  13. 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).

  14. 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.

  15. 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

  16. 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)

  17. 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)

  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. 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

  20. 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)

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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)

  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. 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)

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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)

  19. 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)

  20. 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)

  1. 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.

  2. 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)

  3. 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.

  4. 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)

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.)

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  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. 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

  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. 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

  18. 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)

  19. 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

  20. 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.)

  1. 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

  2. [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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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.)

  8. 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

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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.)

  16. 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

  17. 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)

  18. 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.

  19. 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

  20. 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)

  1. 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

  2. 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...

  3. 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)

  4. 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

  5. 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

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. 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)

  15. 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)

  16. 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.)

  17. 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

  18. 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

  19. 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

  20. 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.)

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. 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.

  13. 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.)

  14. 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.

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.)

  20. 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

  1. 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.)

  2. 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)

  3. 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

  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. 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

  6. 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.

  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. 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.)

  10. 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.)

  11. 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.)

  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. 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)

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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.)

  20. 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)

  1. 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.

  2. 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)

  3. 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)

  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. 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...

  8. 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.

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  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. 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.)

  6. 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).

  7. 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

  8. 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

  9. 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.

  10. 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)