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Sample records for aditya tokamak

  1. Assembly of Aditya upgrade tokamak

    International Nuclear Information System (INIS)

    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)

  2. Radiation power measurement on the ADITYA tokamak

    Science.gov (United States)

    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.

  3. SST and ADITYA tokamak research in India

    International Nuclear Information System (INIS)

    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)

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

    Indian Academy of Sciences (India)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Indian Academy of Sciences (India)

    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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  3. Ion cyclotron resonance heating system on Aditya

    Indian Academy of Sciences (India)

    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.

  4. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meglicki, Z

    1995-09-19

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs.

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

    Indian Academy of Sciences (India)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    Indian Academy of Sciences (India)

    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.

  9. Status of tokamak research

    International Nuclear Information System (INIS)

    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. Tokamak Systems Code

    International Nuclear Information System (INIS)

    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

  11. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

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

  12. Tokamak concept innovations

    International Nuclear Information System (INIS)

    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

  13. Tokamak ARC damage

    International Nuclear Information System (INIS)

    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

  14. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    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)

  15. Research using small tokamaks

    International Nuclear Information System (INIS)

    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

  16. Joint research using small tokamaks

    International Nuclear Information System (INIS)

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

  17. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    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

  18. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    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

  19. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    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

  20. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  1. Advanced tokamak concepts

    NARCIS (Netherlands)

    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

  2. Advanced tokamak concepts

    NARCIS (Netherlands)

    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

  3. Research using small tokamaks

    International Nuclear Information System (INIS)

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

  4. Research using small tokamaks

    International Nuclear Information System (INIS)

    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

  5. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    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

  6. Next tokamak facility

    International Nuclear Information System (INIS)

    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

  7. Tritium catalyzed deuterium tokamaks

    International Nuclear Information System (INIS)

    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)

  8. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    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.)

  9. Tokamak pump limiters

    International Nuclear Information System (INIS)

    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

  10. [High beta tokamak research

    International Nuclear Information System (INIS)

    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

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

    NARCIS (Netherlands)

    Box, F. M. A.; Howard, J.; VandeKolk, E.; Meijer, F. G.

    1997-01-01

    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. S

  12. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  13. Magnetic confinement experiment -- 1: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1994-12-31

    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.

  14. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  16. Tokamak burn control

    International Nuclear Information System (INIS)

    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

  17. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. ITER tokamak device

    Science.gov (United States)

    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.

  19. Edge turbulence in tokamaks

    Science.gov (United States)

    Nedospasov, A. V.

    1992-12-01

    Edge turbulence is of decisive importance for the distribution of particle and energy fluxes to the walls of tokamaks. Despite the availability of extensive experimental data on the turbulence properties, its nature still remains a subject for discussion. This paper contains a review of the most recent theoretical and experimental studies in the field, including mainly the studies to which Wootton (A.J. Wooton, J. Nucl. Mater. 176 & 177 (1990) 77) referred to most in his review at PSI-9 and those published later. The available theoretical models of edge turbulence with volume dissipation due to collisions fail to fully interpret the entire combination of experimental facts. In the scrape-off layer of a tokamak the dissipation prevails due to the flow of current through potential shifts near the surface of limiters of divertor plates. The different origins of turbulence at the edge and in the core plasma due to such dissipation are discussed in this paper. Recent data on the electron temperature fluctuations enabled one to evaluate the electric probe measurements of turbulent flows of particles and heat critically. The latest data on the suppression of turbulence in the case of L-H transitions are given. In doing so, the possibility of exciting current instabilities in biasing experiments (rather than only to the suppression of existing turbulence) is given some attention. Possible objectives of further studies are also discussed.

  20. Dust Measurements in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  1. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  3. Research using small tokamaks

    International Nuclear Information System (INIS)

    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

  4. Spheromak injection into a tokamak

    OpenAIRE

    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...

  5. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    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

  6. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

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

  7. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

    2003-12-01

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

  8. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    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

  10. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    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

  11. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  12. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  13. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

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

  14. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

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

  15. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    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.)

  16. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    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

  17. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    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

  18. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    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

  19. Transport of Dust Particles in Tokamak Devices

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Bootstrap Current in Spherical Tokamaks

    Institute of Scientific and Technical Information of China (English)

    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.

  1. Options for an ignited tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.

    1984-02-01

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

  2. A compact Tokamak transmutation reactor

    Institute of Scientific and Technical Information of China (English)

    QiuLi-Jian; XiaoBing-Jia

    1997-01-01

    The low aspect ration tokamak is proposed for the driver of a transmutation reactor.The main parameters of the reactor core,neutronic analysis of the blanket are given>the neutron wall loading can be lowered from the magnitude order of 1 MW/m2 to 0.5MW/m2 which is much easier to reach in the near future,and the transmutation efficiency (fission/absorption ratio)is raised further.The blanket power density is about 200MW/m3 which is not difficult to deal with.The key components such as diverter and center conductor post are also designed and compared with conventional TOkamak,Finally,by comparison with the other drivers such as FBR,PWR and accelerator,it can be anticipated that the low aspect ratio transmutation reactor would be one way of fusion energy applications in the near future.

  3. Equilibrium Reconstruction in EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    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.

  4. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    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

  5. Magnetic confinement experiment. I: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM`y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nT{tau}`s {approximately} 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.

  6. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    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

  7. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    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

  8. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    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

  9. The disruptive instability in Tokamak plasmas

    NARCIS (Netherlands)

    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

  10. The role of limiter in Egyptor Tokamak

    CERN Document Server

    Ei-Sisi, A B

    2002-01-01

    In Egyptor Tokamak, the limiter is used for separation of the plasma from the vessel. In this work an overview of limiter types, and construction of limiter in Egyptor Tokamak is discussed. Also simulation results of the radial electron density distribution in case of limiter are presented. The results of the simulation are in agreement with the experimental and analytical results.

  11. Breakdown in the pretext tokamak

    International Nuclear Information System (INIS)

    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

  12. Atomic physics in tokamak plasmas

    International Nuclear Information System (INIS)

    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.)

  13. Control of a burning tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    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.)

  15. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

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

  16. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    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. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  18. D-D tokamak reactor studies

    International Nuclear Information System (INIS)

    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

  19. Simulating Plasma Turbulence in Tokamaks

    CERN Document Server

    Kepner, J V; Decyk, V; Kepner, Jeremy; Parker, Scott; Decyk, Viktor

    1997-01-01

    A challenging and fundamental research problem is the better understanding and control of the turbulent transport of heat in present-day tokamak fusion experiments. Recent developments in numerical methods along with enormous gains in computing power have made large-scale simulations an important tool for improving our understanding of this phenomena. Simulating this highly non-linear behavior requires solving for the perturbations of the phase space distribution function in five dimensions. We use a particle-in-cell approach to solve the equations. The code has been parallelized for a variety of architectures (C90, CM-5, T3D) using a 1-D domain decomposition along the toroidal axis, for which the number of particles in each cell remains approximately constant. The quasi-uniform distribution of particles, which minimizes load imbalance, coupled with the relatively small movement of particles across cells, which minimizes communications, makes this problem ideally suited to massively parallel architectures. We...

  20. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    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

  1. The JT-60 tokamak machine

    International Nuclear Information System (INIS)

    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.)

  2. Characteristics of Plasma Turbulence in the Mega Amp Spherical Tokamak

    CERN Document Server

    Ghim, Young-chul

    2013-01-01

    Turbulence is a major factor limiting the achievement of better tokamak performance as it enhances the transport of particles, momentum and heat which hinders the foremost objective of tokamaks. Hence, understanding and possibly being able to control turbulence in tokamaks is of paramount importance, not to mention our intellectual curiosity of it.

  3. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

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

  4. Electron thermal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    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)

  5. Simulation of burning tokamak plasmas

    International Nuclear Information System (INIS)

    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.)

  6. Microtearing modes in tokamak discharges

    Science.gov (United States)

    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.

  7. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

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

  8. Mass spectrometry instrumentation in TN (Novillo Tokamak)

    International Nuclear Information System (INIS)

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

  9. The Spherical Tokamak MEDUSA for Mexico

    Science.gov (United States)

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

    2011-10-01

    The former spherical tokamak MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

  10. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-12-01

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

  11. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    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.

  12. Electron cyclotron emission diagnostics on KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. A method for tokamak neutronics calculations

    International Nuclear Information System (INIS)

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

  14. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    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

  15. Multichannel submillimeter interferometer for tokamak density measurements

    International Nuclear Information System (INIS)

    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

  16. Can better modelling improve tokamak control?

    International Nuclear Information System (INIS)

    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

  17. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    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

  18. Electronic system of TBR tokamak device

    International Nuclear Information System (INIS)

    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)

  19. Spontaneous generation of rotation in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    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

  1. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. UCLA Tokamak Program Close Out Report.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Toroidal Alfven wave stability in ignited tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    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

  5. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    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

  6. Radioactivity evaluation for the KSTAR tokamak

    International Nuclear Information System (INIS)

    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)

  7. Analysis of sawtooth relaxation oscillations in tokamaks

    International Nuclear Information System (INIS)

    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

  8. Tokamak Transport Studies Using Perturbation Analysis

    NARCIS (Netherlands)

    Cardozo, N. J. L.; Dehaas, J. C. M.; Hogeweij, G. M. D.; Orourke, J.; Sips, A.C.C.; Tubbing, B. J. D.

    1990-01-01

    Studies of the transport properties of tokamak plasmas using perturbation analysis are discussed. The focus is on experiments with not too large perturbations, such as sawtooth induced heat and density pulse propagation, power modulation and oscillatory gas-puff experiments. The approximations made

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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

  11. Banana orbits in elliptic tokamaks with hole currents

    Science.gov (United States)

    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.

  12. Systems studies of high-field tokamak ignition experiments

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  14. Numerical studies of edge localized instabilities in tokamaks

    International Nuclear Information System (INIS)

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

  15. KTM Tokamak operation scenarios software infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, V.; Baystrukov, K.; Golobkov, YU.; Ovchinnikov, A.; Meaentsev, A.; Merkulov, S.; Lee, A. [National Research Tomsk Polytechnic University, Tomsk (Russian Federation); Tazhibayeva, I.; Shapovalov, G. [National Nuclear Center (NNC), Kurchatov (Kazakhstan)

    2014-10-15

    One of the largest problems for tokamak devices such as Kazakhstan Tokamak for Material Testing (KTM) is the operation scenarios' development and execution. Operation scenarios may be varied often, so a convenient hardware and software solution is required for scenario management and execution. Dozens of diagnostic and control subsystems with numerous configuration settings may be used in an experiment, so it is required to automate the subsystem configuration process to coordinate changes of the related settings and to prevent errors. Most of the diagnostic and control subsystems software at KTM was unified using an extra software layer, describing the hardware abstraction interface. The experiment sequence was described using a command language. The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.

  16. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1996-06-01

    A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).

  17. Boundary Plasma Turbulence Simulations for Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  18. The Spherical Tokamak MEDUSA for Costa Rica

    Science.gov (United States)

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

    2012-10-01

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

  19. Module description of TOKAMAK equilibrium code MEUDAS

    International Nuclear Information System (INIS)

    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)

  20. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-01-01

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

  2. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

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

  3. EU Integrated Tokamak Modelling (ITM) Task Force

    Institute of Scientific and Technical Information of China (English)

    A Becoulet

    2007-01-01

    @@ At the end of 2003, the European Fusion Development Agreement (EFDA) structure set-up a long-term European task force (TF) in charge of "co-ordinating the development of a coherent set of validated simulation tools for the purpose of benchmarking on existing tokamak experiments, with the ultimate aim of providing a comprehensive simulation package for ITER plasmas" [http://www.efda-taskforce-itm.org/].

  4. Smaller coil systems for tokamak reactors

    International Nuclear Information System (INIS)

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

  5. Resistive interchange instability in reversed shear tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Furukawa, Masaru; Nakamura, Yuji; Wakatani, Masahiro [Graduate School of Energy Science, Kyoto University, Uji, Kyoto (Japan)

    1999-04-01

    Resistive interchange modes become unstable due to the magnetic shear reversal in tokamaks. In the present paper, the parameter dependences, such as q (safety factor) profile and the magnetic surface shape are clarified for improving the stability, using the local stability criterion. It is shown that a significant reduction of the beta limit is obtained for the JT-60U reversed shear configuration with internal transport barrier, since the local pressure gradient increases. (author)

  6. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

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

  7. Self-Organized Stationary States of Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

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

  8. Self-Organized Stationary States of Tokamaks.

    Science.gov (United States)

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

    2015-11-20

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

  9. Microtearing modes and anomalous transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Drake, J.F.; Gladd, N.T.; Liu, C.S.; Chang, C.L.

    1980-04-14

    Microtearing (high-m) modes driven by the electron temperature gradient are found to be unstable for present tokamak parameters. A self-consistent calculation of the nonlinear saturation of this instability yields magnetic fluctuations vertical-barBvertical-bar/B approx. = rho/sub e//L/sub T/. The associated crossfield electron thermal conductivity is shown to be inversely proportional to density, consistent with Alcator scaling, and comparable in magnitude with that inferred from experiments.

  10. Neoclassical transport in high [beta] tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high [beta] large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high [beta] large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low [beta] values by a factor ([var epsilon]/q[sup 2][beta])[sup [1/2

  11. ECH on the MTX [Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

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

  12. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

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

    2015-11-01

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

  13. SOL Width Scaling in the MAST Tokamak

    Science.gov (United States)

    Ahn, Joon-Wook; Counsell, Glenn; Connor, Jack; Kirk, Andrew

    2002-11-01

    Target heat loads are determined in large part by the upstream SOL heat flux width, Δ_h. Considerable effort has been made in the past to develop analytical and empirical scalings for Δh to allow reliable estimates to be made for the next-step device. The development of scalings for a large spherical tokamak (ST) such as MAST is particularly important both for development of the ST concept and for improving the robustness of scalings derived for conventional tokamaks. A first such scaling has been developed in MAST DND plasmas. The scaling was developed by flux-mapping data from the target Langmuir probe arrays to the mid-plane and fitting to key upstream parameters such as P_SOL, bar ne and q_95. In order to minimise the effects of co-linearity, dedicated campaigns were undertaken to explore the widest possible range of each parameter while keeping the remainder as fixed as possible. Initial results indicate a weak inverse dependence on P_SOL and approximately linear dependence on bar n_e. Scalings derived from consideration of theoretical edge transport models and integration with data from conventional devices is under way. The established scaling laws could be used for the extrapolations to the future machine such as Spherical Tokamak Power Plant (STPP). This work is jointly funded by Euratom and UK Department of Trade and Industry. J-W. Ahn would like to recognise the support of a grant from the British Foreign & Commonwealth Office.

  14. Relativistic runaway electrons in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Jaspers, R.E.

    1995-02-03

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

  15. Deposit of thin films for Tokamaks conditioning

    International Nuclear Information System (INIS)

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

  16. Soft-X-Ray Tomography Diagnostic at the Rtp Tokamak

    NARCIS (Netherlands)

    Da Cruz, D. F.; Donne, A. J. H.

    1994-01-01

    An 80-channel soft x-ray tomography system has been constructed for diagnosing the RTP (Rijnhuizen Tokamak Project) tokamak plasma. Five pinhole cameras, each with arrays of 16 detectors are distributed more or less homogeneously around a poloidal plasma cross section. The cameras are positioned clo

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    Haverkort, J.W.

    2013-01-01

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

  1. Experimental studies of tokamak plasma in IPP Prague

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Lower hybrid heating experiments in tokamaks: an overview

    International Nuclear Information System (INIS)

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

  4. A simulation study of a controlled tokamak plasma

    Science.gov (United States)

    Fujii, N.; Niwa, Y.

    1980-03-01

    A tokamak circuit theory, including results of numerical simulation studies, is applied to a control system synthesized for a Joule heated tokamak plasma. The treatment is similar to that of Ogata and Ninomiya (1979) except that in this case a quadrupole field coil current is considered coexisting with image induced on a vacuum chamber.

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

    International Nuclear Information System (INIS)

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

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

    NARCIS (Netherlands)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ignat, D.W.

    1987-01-01

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

  8. Experimental data base of Tokamak KTM physical diagnostics

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Indian Academy of Sciences (India)

    R Kumar; S K Saha

    2000-11-01

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

  11. The Experiments of the small Spherical Tokamak Gutta

    International Nuclear Information System (INIS)

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

  12. MHD stability of advanced tokamak scenarios

    International Nuclear Information System (INIS)

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

  13. Industry roles in the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

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

  14. Differential and Integral Models of TOKAMAK

    Directory of Open Access Journals (Sweden)

    Ivo Dolezel

    2004-01-01

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

  15. Scaling studies of beam-heated tokamaks

    International Nuclear Information System (INIS)

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

  16. Spherical tokamak research for fusion reactor

    International Nuclear Information System (INIS)

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

  17. Iron forbidden lines in tokamak discharges

    International Nuclear Information System (INIS)

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

  18. Profile control for an alternative spherical tokamak

    International Nuclear Information System (INIS)

    Magnetically driven plasma guns that are inserted around a flux conserver at definite angular intervals are considered. The creation and the control of plasma channels are examined. By means of the hybrid model developed, both a system analysis of the Alternative Spherical Tokamak (AST) and relevant computational experiments have been carried out. In addition, by using the results obtained from the numerical scheme, the complex non-inductive current drive mechanisms of bootstrap and helicity injection in the AST system are discussed in detail. (author). 2 refs, 2 figs

  19. DIII-D Advanced Tokamak Research Overview

    International Nuclear Information System (INIS)

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

  20. Application of MDSplus on EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

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

    2007-01-01

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

  1. Bolometer measurement on HT-6B tokamak

    International Nuclear Information System (INIS)

    This paper discribes the structure, methods of calibration and measurement system of a metal foil resistor bolometer which is developed for measuring the radiation power of high temperature plasmas. The radiation loss and neutral flux loss in HT-6B tokamak have been measured by using the bolometer. The following results were obtained: (1) A large, nearly constant fraction (∼50%) of the input power was lost to the wall by radiation and energetic neutrals during the quasisteady phase of a normal discharges; (2) The power loss linearly increased with the discharge current Ip; (3) During disruption, most of the plasma energy was lost by radiation and neutrals

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  3. Tokamak advanced pump limiter experiments and analysis

    International Nuclear Information System (INIS)

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

  4. Magnetic diagnostics for the lithium tokamak experiment.

    Science.gov (United States)

    Berzak, L; Kaita, R; Kozub, T; Majeski, R; Zakharov, L

    2008-10-01

    The lithium tokamak experiment (LTX) is a spherical tokamak with R(0)=0.4 m, a=0.26 m, B(TF) approximately 3.4 kG, I(P) approximately 400 kA, and pulse length approximately 0.25 s. The focus of LTX is to investigate the novel low-recycling lithium wall operating regime for magnetically confined plasmas. This regime is reached by placing an in-vessel shell conformal to the plasma last closed flux surface. The shell is heated and then coated with liquid lithium. An extensive array of magnetic diagnostics is available to characterize the experiment, including 80 Mirnov coils (single and double axis, internal and external to the shell), 34 flux loops, 3 Rogowskii coils, and a diamagnetic loop. Diagnostics are specifically located to account for the presence of a secondary conducting surface and engineered to withstand both high temperatures and incidental contact with liquid lithium. The diagnostic set is therefore fabricated from robust materials with heat and lithium resistance and is designed for electrical isolation from the shell and to provide the data required for highly constrained equilibrium reconstructions. PMID:19044600

  5. Compact ignition tokamak physics and engineering basis

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is a high-field, compact tokamak design whose objective is the study of physics issues associated with burning plasmas. The toroidal and poloidal field coils employ a copper-steel laminate, manufactured by explosive-bonding techniques, to support the forces generated by the design fields: 10 T toroidal field at the plasma center; 21 T in the OH solenoid. A combination of internal and external PF coils provides control of the equilibrium and the ability to sweep the magnetic separatrix across the divertor plates during a pulse. At temperatures and βα levels characteristic of ITER designs, the fusion power in CIT approaches 800 MW and can be the limiting factor in the pulse length. Ignition requires that the confinement time exceed present L-mode scalings by about a factor of two, which is anticipated to occur as a result of the operational flexibility incorporated into the design. Conventional operating limits given by 20 e and qψ ≤ 3.2 have been chosen and, in the case of MHD limits, have been justified by ideal stability analysis. The power required for CIT ignition ranges from 10 MW to 40 MW or more, depending on confinement assumptions, and either ICRF or ECRF heating, or both, will be used. (author). 17 refs, 6 figs, 1 tab

  6. Physics evaluation of compact tokamak ignition experiments

    International Nuclear Information System (INIS)

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

  7. Conceptual tokamak design at high neutron fluence

    International Nuclear Information System (INIS)

    For the future fusion reactor, it is important to design an experimental device that can be performed testing in-vessel components including tritium breeding modules relevant to the future fusion reactor with high neutron fluence. To realize this requirement, a conceptual tokamak design has been performed in accordance with plasma performance and shape at quasi-steady-state operation. One of the promising scenarios for this purpose is proposed to produce the plasma at the outward shifted radial position with a small minor radius for reasonable plasma parameters. From the analytical results, an appropriate space can be found for neutron shielding so that additional neutron shielding can be installed to protect the tokamak components from any neutron damages under the neutron fluence of 1 MWa m-2. Based on the structural analyses, a two-stage blanket module concept is proposed, i.e. one shielding block with the first wall assembly during high Q operation and two shielding blocks or additional tritium breeding modules during quasi-steady state operation

  8. Electromagnetic simulations of tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Michael; Mishchenko, Alexey [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Assoziation, Wendelsteinstrasse 1, 17491 Greifswald (Germany)

    2014-07-01

    A practical fusion reactor will require a plasma β of around 5%. In this range Alfvenic effects become important. Since a practical reactor will also produce energetic alpha particles, the interaction between Alfvenic instabilities and fast ions is of particular interest. We have developed a fluid electron, kinetic ion hybrid model that can be used to study this problem. Compared to fully gyrokinetic electromagnetic codes, hybrid codes offer faster running times and greater flexibility, at the cost of reduced completeness. The model has been successfully verified against the worldwide ITPA Toroidal Alfven Eigenmode (TAE) benchmark, and the ideal MHD code CKA for the internal kink mode in a tokamak. Use of the model can now be turned toward cases of practical relevance. Current work focuses on simulating fishbones in a tokamak geometry, which may be of relevance to ITER, and producing the first non-perturbative self-consistent simulations of TAE in a stellarator, which may be of relevance both to Wendelstein 7-X and any future stellarator reactor. Preliminary results of these studies are presented.

  9. Ion cyclotron system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-01

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

  10. Ion cyclotron system design for KSTAR tokamak

    International Nuclear Information System (INIS)

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

  11. System studies of compact ignition tokamaks

    International Nuclear Information System (INIS)

    The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B2a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations

  12. Boronization of Russian tokamaks from carborane precursors

    International Nuclear Information System (INIS)

    A new and cheap boronization technique using the nontoxic and nonexplosive solid substance carborane has been developed and successfully applied to the Russian tokamaks T-11M, T-3M, T-10 and TUMAN-3. The glow discharge in a mixture of He and carborane vapor produced the amorphous B/C coating with the B/C ratio varied from 2.0-3.7. The deposition rate was about 150 nm/h. The primary effect of boronization was a significant reduction of the impurity influx and the plasma impurity contamination, a sharp decrease of the plasma radiated power, and a decrease of the effective charge. Boronization strongly suppressed the impurity influx caused by additional plasma heating. ECR- and ICR-heating as well as ECR current drive were more effective in boronized vessels. Boronization resulted in a significant extension of the Ne- and q-region of stable tokamak operation. The density limit rose strongly. In Ohmic H-mode energy confinement time increased significantly (by a factor of 2) after boronization. It rose linearly with plasma current Ip and was 10 times higher than Neo-Alcator time at maximum current. ((orig.))

  13. Thomson scattering on the PRETEXT Tokamak

    International Nuclear Information System (INIS)

    Ruby laser Thomson scattering was performed on the PRETEXT tokamak. A 10 Joule Q-switched laser and a 1 meter 10 channel polychromator were used to diagnose the electron temperature and density profiles in the PRETEXT plasma. These parameters were measured as a function of time and radial position on a shot to shot basis. The density measurement was calibrated by Rayleigh and Raman scattering and by comparison with data from a 4 mm microwave interferometer. Electron densities ranging from 1 x 1012 cm-3 to 2 x 1013 cm-3 and temperatures ranging from 3 eV to 400 eV were observed. Detailed measurements were made throughout the 40 ms discharge with particular emphasis on the current rise phase. The Thomson scattering data was used as input to a one dimensional magnetic diffusion code. This code modelled the evolution of the current density and safety factor profiles. The results of this analysis were compared with existing theories of tokamak current penetration. The growth of resitive MHD tearing modes was proposed as a likely explanation for the anomalously rapid current penetration observed in PRETEXT

  14. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

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

  15. System studies of compact ignition tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Galambos, J.D.; Blackfield, D.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Selcow, E.

    1987-08-01

    The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B/sup 2/a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations.

  16. Alfven wave studies on a tokamak

    International Nuclear Information System (INIS)

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

  17. Anomalous transport in the tokamak edge

    International Nuclear Information System (INIS)

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

  18. Equilibrium reconstruction in the START tokamak

    Science.gov (United States)

    Appel, L. C.; Bevir, M. K.; Walsh, M. J.

    2001-02-01

    The computation of magnetic equilibria in the START spherical tokamak is more difficult than those in more conventional large aspect ratio tokamaks. This difficulty arises partly as a result of the use of induction compression to generate high current plasma, as this precludes the positioning of magnetic diagnostics close to the outboard side of the plasma. In addition, the effect of a conducting wall with a high, but finite, conductivity must be included. A method is presented for obtaining plasma equilibrium reconstructions based on the EFIT code. New constraints are used to relate isoflux surface locations deduced from radial profile measurements of electron temperature. A model of flux diffusion through the vessel wall is developed. It is shown that neglecting flux diffusion in the vessel wall can lead to a significant underestimate in the calculation of the plasma βt. Using a relatively sparse set of magnetic signals, βt can be obtained to within a fractional error of +/-10%. Using constraints to relate isoflux surface locations, the principle involved in determining the internal q profile is demonstrated.

  19. Soft x-ray tomography on tokamaks using flux coordinates

    International Nuclear Information System (INIS)

    Methods of inverting line integrated data using coordinates, which are adapted to problems arising from Hamiltonian flows, are presented. They are exemplified for measurements of soft x-rays on tokamaks with widely arbitrary poloidal cross section. Boundary conditions can be met and cause fewer 'ghosts' for most of the present day tokamaks. The soft x-ray measurements are then used to improve the flux function Ψ as obtained from codes using magnetic measurements as input. We investigate oscillatory phenomena such as sawtooth crash precursors on tokamaks by decomposing the profiles into space-like eigenfunctions and their time dependencies. (author)

  20. Ions Measurement at the Edge of HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    Ling Bili; Wang Enyao; Gao wei; Wan Baonian; Li Jiangang

    2005-01-01

    A reliable method of measuring ions and ion temperature in tokamak plasma is necessary, for which an omegatron-like instrument has been developed on the HT-7 tokamak. The basic layout of the omegatron-like instrument is shown in this article. The measurement of working gas ion has been performed in the last experimental campaign on HT-7 tokamak. The relations among ion current, the electron repeller voltage and trap voltage have been investigated. This omegatron-like instrument has also provided the edge-plasma ion temperature.

  1. A systems assessment of the five Starlite tokamak power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.

    1996-07-01

    The ARIES team has assessed the power-plant attractiveness of the following five tokamak physics regimes: (1) steady state, first stability regime; (2) pulsed, first stability regime; (3) steady state, second stability regime; (4) steady state, reversed shear; and (5) steady state, low aspect ratio. Cost-based systems analysis of these five tokamak physics regimes suggests that an electric power plant based upon a reversed-shear tokamak is significantly more economical than one based on any of the other four physics regimes. Details of this comparative systems analysis are described herein.

  2. Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

    Science.gov (United States)

    Zheng, Linjin; Horton, W.; Miura, H.; Shi, T. H.; Wang, H. Q.

    2016-08-01

    Owing to the large ion orbits, excessive electrons can accumulate at tokamak edge. We find that the nonneutralized electrons at tokamak edge can contribute an electric compressive stress in the direction parallel to magnetic field by their mutual repulsive force. By extending the Chew-Goldburger-Low theory (Chew et al., 1956 [13]), it is shown that this newly recognized compressive stress can significantly change the plasma average magnetic well, so that a stabilization of magnetohydrodynamic modes in the pedestal can result. This linear stability regime helps to explain why in certain parameter regimes the tokamak high confinement can be rather quiet as observed experimentally.

  3. Tokamak Plasmas : Observation of floating potential asymmetry in the edge plasma of the SINP tokamak

    Indian Academy of Sciences (India)

    Krishnendu Bhattacharyya; N R Ray

    2000-11-01

    Edge plasma properties in a tokamak is an interesting subject of study from the view point of confinement and stability of tokamak plasma. The edge plasma of SINP-tokamak has been investigated using specially designed Langmuir probes. We have observed a poloidal asymmetry of floating potentials, particularly the top-bottom floating potential differences are quite noticeable, which in turn produces a vertical electric field (v). This v remains throughout the discharge but changes its direction at certain point of time which seems to depend on applied vertical magnetic field v).

  4. Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR

    Energy Technology Data Exchange (ETDEWEB)

    Sa, Wanderley Pires de

    1996-12-31

    The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.

  5. Development of tokamak reactor system analysis code NEW-TORSAC

    Science.gov (United States)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  6. Engineering development aspects of the HL-2A tokamak

    International Nuclear Information System (INIS)

    The HL-2A tokamak (design values: major radius 1.65 m, minor radius 0.4 m, plasma current 0.48 MA and toroidal field 2.8 T) is the first tokamak with an operating divertor in China. It is characterized by a large closed divertor chamber. This unique feature will make significant contributions to enhance our understanding of complex divertor plasma physics and to help validating divertor physics modelings. The engineering design, development, testing and commissioning of the HL-2A tokamak are described in this paper. Preliminary results show that the HL-2A tokamak has been successfully operated in the divertor configuration. The major parameters: plasma current Ip=168 kA, toroidal field BT=1.4 T, plasma line average density ne=1.7 x 1019 m-3, limiting vacuum pv=4.6 x 10-6 Pa, were achieved at the end of 2003. (authors)

  7. HYFIRE: a tokamak- high-temperature electrolysis system

    International Nuclear Information System (INIS)

    Brookhaven National Laboratory is involved in a conceptual design study of a commercial nuclear power system which utilizes high-temperature electrolysis to produce synthetic fuels. The system is called HYFIRE. It includes a tokamak fusion power reactor supplying electrical and thermal energy to an array of electrolytes. The electrolytes produce hydrogen which can be used either directly as a fuel or in the production of hydrocarbons. The purpose of the study is to provide a mechanism for DOE to further assess the commercial potential of fusion using a tokamak reactor to produce synthetic fuel. The HYFIRE design is based on the tokamak commercial power reactor, STARFIRE. STARFIRE uses the deuterium/tritium/lithium fuel cycle. The HYFIRE study assumes the plasma shape and characteristics of STARFIRE study but uses a different blanket design. This study is particularly interested in the possibility of using the STARFIRE tokamak in the production of synthetic fuels

  8. Toroidicity Dependence of Tokamak Edge Safety Factor and Shear

    Institute of Scientific and Technical Information of China (English)

    SHIBingren

    2002-01-01

    In large tokamak device and reactor designs, the relationship between the toroidal current and the edge safety factor is very important because this will determine the eventual device or reactor size according to MHD stability requirements. In many preliminary

  9. Compact Ignition Tokamak Program: status of FEDC studies

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, C.A.

    1985-01-01

    Viewgraphs on the Compact Ignition Tokamak Program comprise the report. The technical areas discussed are the mechanical configuration status, magnet analysis, stress analysis, cooling between burns, TF coil joint, and facility/device layout options. (WRF)

  10. NEOCLASSICAL TRANSPORT IN A TOKAMAK WITH ELECTRIC SHEAR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Neoclassical transport theory for a tokamak in the presence of a large radial electric field with shear is developed using Hamiltonian formalism. Diffusion coefficients are derived in both the plateau and banana regimes where the squeezing factor in coefficients can greatly affect diffusion at the plasma edge. Rotation speeds are calculated in the scrape-off region. They are in good agreement with the measurements on the TdeV tokamak.

  11. An emerging understanding of H-mode discharges in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Groebner, R.J.

    1992-12-01

    A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the [upsilon][sub E][sup [yields

  12. Technique for plasma filament stabilization in a tokamak

    International Nuclear Information System (INIS)

    The invention is related to the field of automatic control of thermonuclear device processes and can be used in control systems of plasma filament stabilization by large radius in tokamak type thermolnuclear devices. The economic effect of the suggested technique is caused by improvement of stabilization of optimum (from the viewpoint of the decrease of plasma energy losses) plasma filament position in the tokamak-reactor which results in the decrease of power of additional plasma heating systems

  13. Design of a microwave calorimeter for the microwave tokamak experiment

    Energy Technology Data Exchange (ETDEWEB)

    Marinak, M. (California Univ., Berkeley, CA (USA))

    1988-10-07

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs.

  14. Development of tokamak reactor systems analysis code 'TORSAC'

    International Nuclear Information System (INIS)

    This report describes Tokamak Reactor Systems Analysis Code ''TORSAC'' which has been developed in order to assess the impact of the design choises on reactor systems and to improve tokamak designs in wide parameter range. This computer code has following functions. (1) Systematic sensitivity analysis for a set of given design parameters, (2) Cost calculation of a new reactor concept designed automatically as a result of systematic sensitivity analysis. (author)

  15. Design of a microwave calorimeter for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs

  16. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-07-01

    Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The

  17. Cooldown of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs

  18. Beam-induced tensor pressure tokamak equilibria

    International Nuclear Information System (INIS)

    D-shaped tensor pressure tokamak equilibria induced by neutral-beam injection are computed. The beam pressure components are evaluated from the moments of a distribution function that is a solution of the Fokker-Planck equation in which the pitch-angle scattering operator is ignored. The level-psub(perpendicular) contours undergo a significant shift away from the outer edge of the device with respect to the flux surfaces for perpendicular beam injection into broad-pressure-profile equilibria. The psub(parallel) contours undergo a somewhat smaller inward shift with respect to the flux surfaces for both parallel and perpendicular injection into broad-pressure-profile equilibria. For peaked-pressure-profile equilibria, the level pressure contours nearly co-incide with the flux surfaces. (author)

  19. Cooldown of the Compact Ignition Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Keeton, D.C.

    1987-08-01

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.

  20. Passive runaway electron suppression in tokamak disruptions

    International Nuclear Information System (INIS)

    Runaway electrons created in disruptions pose a serious problem for tokamaks with large current. It would be desirable to have a runaway electron suppression method which is passive, i.e., a method that does not rely on an uncertain disruption prediction system. One option is to let the large electric field inherent in the disruption drive helical currents in the wall. This would create ergodic regions in the plasma and increase the runaway losses. Whether these regions appear at a suitable time and place to affect the formation of the runaway beam depends on disruption parameters, such as electron temperature and density. We find that it is difficult to ergodize the central plasma before a beam of runaway current has formed. However, the ergodic outer region will make the Ohmic current profile contract, which can lead to instabilities that yield large runaway electron losses

  1. Ignition probabilities for Compact Ignition Tokamak designs

    International Nuclear Information System (INIS)

    A global power balance code employing Monte Carlo techniques had been developed to study the ''probability of ignition'' and has been applied to several different configurations of the Compact Ignition Tokamak (CIT). Probability distributions for the critical physics parameters in the code were estimated using existing experimental data. This included a statistical evaluation of the uncertainty in extrapolating the energy confinement time. A substantial probability of ignition is predicted for CIT if peaked density profiles can be achieved or if one of the two higher plasma current configurations is employed. In other cases, values of the energy multiplication factor Q of order 10 are generally obtained. The Ignitor-U and ARIES designs are also examined briefly. Comparisons of our empirically based confinement assumptions with two theory-based transport models yield conflicting results. 41 refs., 11 figs

  2. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  3. Sliding Mode Control of a Tokamak Transformer

    Energy Technology Data Exchange (ETDEWEB)

    Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.

    2012-06-08

    A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.

  4. Low Z impurity transport in tokamaks

    International Nuclear Information System (INIS)

    Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed

  5. Instrumentation and controls of an ignited tokamak

    International Nuclear Information System (INIS)

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

  6. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  7. Safety factor profile control in a tokamak

    CERN Document Server

    Bribiesca Argomedo, Federico; Prieur, Christophe

    2014-01-01

    Control of the Safety Factor Profile in a Tokamak uses Lyapunov techniques to address a challenging problem for which even the simplest physically relevant models are represented by nonlinear, time-dependent, partial differential equations (PDEs). This is because of the  spatiotemporal dynamics of transport phenomena (magnetic flux, heat, densities, etc.) in the anisotropic plasma medium. Robustness considerations are ubiquitous in the analysis and control design since direct measurements on the magnetic flux are impossible (its estimation relies on virtual sensors) and large uncertainties remain in the coupling between the plasma particles and the radio-frequency waves (distributed inputs). The Brief begins with a presentation of the reference dynamical model and continues by developing a Lyapunov function for the discretized system (in a polytopic linear-parameter-varying formulation). The limitations of this finite-dimensional approach motivate new developments in the infinite-dimensional framework. The t...

  8. The Procedure for Assembling the EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    Wu Songtao

    2005-01-01

    Due to the complicated constitution and high precision requirements of the EAST superconducting tokamak, a meticulous assembling procedure and measurement scheme must be established. The big size and mass of the EAST machine's components and complicated configuration with tight installation tolerances call for a highly careful assembling procedure. The assembling procedure consists of three main sub-procedures for the assembling of the base, of the tori of the VV, the vacuum vessel TS and the TF, and of the peripheral parts respectively. Before the assembly, a reference framework has been set up by means of an industrial measurement system with reference fiducial targets fixed on the wall of the test hall. In this paper, the assembling procedure is described in detail, the survey control system of the assembly is discussed, and progress in the assembly work is also reported.

  9. Natural Fueling of a Tokamak Fusion Reactor

    CERN Document Server

    Wan, Weigang; Chen, Yang; Perkins, Francis W

    2009-01-01

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

  10. Interactions of tokamak plasma with solid walls

    International Nuclear Information System (INIS)

    The interactions of tokamak fusion plasmas with solid walls of the devices were investigated on special model systems. The elastic recoil detection method was used for the determination of absolute hydrogen concentration. For the calibration of the method the scattering cross sections were measured in large ranges of scattering angle and energy. The erosion and deformation of wall surfaces were investigated by reemission of accelerated He ions. Theoretical models were developed to describe the surface undulation discovered earlier, caused by large dose He irradiation. The surface sputtering and segregation were investigated by nuclear methods and the mechanism of sputtering was simulated by computer. The surface deformation and gas reemission of Al surfaces were analyzed by Ar implementation and heat treatment. (D.Gy.) 6 figs

  11. Real time analysis of tokamak discharge parameters

    International Nuclear Information System (INIS)

    The techniques used in implementing two applications of real time analysis of data from the DIII-D tokamak are described. These tasks, which are demanding in both the speed of data acquisition and the speed of computation, execute on hardware capable of acquiring 40 million data samples per second and executing 80 million floating point operations per second. In the first case, a feedback control algorithm executing at a 10 kHz cycle frequency is used to specify the current in the poloidal field coils in order to control the discharge shape. In the second, fast Fourier transforms of Mirnov probe data are used to find the amplitude and frequency of each of eight toroidal mode numbers as a function of time during the discharge. Data sampled continuously at 500 kHz are used to produce results at 2 msec intervals

  12. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

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

  13. Pumped limiter results on TFR Tokamak plasma

    International Nuclear Information System (INIS)

    Pump limiter experiments are carried out in the TFR Tokamak. The pump limiter is located in the outer part of the torus, its double- throat head is made of graphite tiles and it is pumped by a 2000 ls-1 titanium sublimation pump. The first attempts showed that the exhaust efficiency of this pump limiter was low (ε = 1.5% of the total plasma particle efflux). To improve these results, a new limiter head with a single longer throat has been built; particles were better trapped and the pumping provided an important decrease of the recycling coefficient. Geometric features mainly explain the increase by a factor 3.5 of the exhaust efficiency (ε = 5.5%). Ion temperature of the order of a few eV has been deduced from Doppler broadening measurements at the neutralizer plate of the pump limiter

  14. Next tokamak design. Swimming pool type

    International Nuclear Information System (INIS)

    In order to relieve the difficulties of repair and maintenance and to make the reactor size compact, a concept of swimming pool type reactor which is installed in a waterpool has been proposed. A design study of the concept as the Next Tokamak has been carried out with the following major parameters. The reactor has a double null poloidal divertor and blanket with tritium breeding ratio of >1.0, fusion power 420 MW, major radius 5.3 m, plasma radius 1.1 m, Bt on axis 5.2 T, plasma current 3.9 MA. The design study covers the reactor overall systems including reactor structure, reactor cooling system, repair and maintenance, reactor building, etc. As the result of this study the following conclusions were reached. The advantages over a conventional tokamak reactor are as follows: (1) The size of TF coil can be considerably reduced while retaining sufficient space for repair and maintenance because a solid shield is eliminated. (2) Since the distances between plasma and PF coils become small, the required capacity of electric power supply is reduced. (3) Technologies for the repair and maintenance are simplified and disassembling and reassembling of vacuum vessel can be done with realistic and credible remote handling technique. (4) The problem caused by radiation streaming can be considerably eased. (5) Radioactive waste disposal is reduced considerably because a solid shield is eliminated. (6) Because a vacuum vessel may be easily replaced in this concept, it will have a convenient flexibility for an experimental reactor. (7) Advantages of this concept can be also applied to a power reactor. Recently we started a new design of SPTR with slightly modified plasma parameters aiming for smaller-size reactor. In this paper the new design will be discussed briefly. (author)

  15. Ion cyclotron emission in tokamak plasmas

    International Nuclear Information System (INIS)

    Detection of α(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, α particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. α particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central α density in a reactor. (author)

  16. Embedded data acquisition system with MDSPlus

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-15

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

  17. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-07-01

    Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The

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

    International Nuclear Information System (INIS)

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

  19. Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

    International Nuclear Information System (INIS)

    SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

  20. Basic Physics of Tokamak Transport Final Technical Report.

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Amiya K.

    2014-05-12

    The goal of this grant has been to study the basic physics of various sources of anomalous transport in tokamaks. Anomalous transport in tokamaks continues to be one of the major problems in magnetic fusion research. As a tokamak is not a physics device by design, direct experimental observation and identification of the instabilities responsible for transport, as well as physics studies of the transport in tokamaks, have been difficult and of limited value. It is noted that direct experimental observation, identification and physics study of microinstabilities including ITG, ETG, and trapped electron/ion modes in tokamaks has been very difficult and nearly impossible. The primary reasons are co-existence of many instabilities, their broadband fluctuation spectra, lack of flexibility for parameter scans and absence of good local diagnostics. This has motivated us to study the suspected tokamak instabilities and their transport consequences in a simpler, steady state Columbia Linear Machine (CLM) with collisionless plasma and the flexibility of wide parameter variations. Earlier work as part of this grant was focused on both ITG turbulence, widely believed to be a primary source of ion thermal transport in tokamaks, and the effects of isotope scaling on transport levels. Prior work from our research team has produced and definitively identified both the slab and toroidal branches of this instability and determined the physics criteria for their existence. All the experimentally observed linear physics corroborate well with theoretical predictions. However, one of the large areas of research dealt with turbulent transport results that indicate some significant differences between our experimental results and most theoretical predictions. Latter years of this proposal were focused on anomalous electron transport with a special focus on ETG. There are several advanced tokamak scenarios with internal transport barriers (ITB), when the ion transport is reduced to

  1. Importance of effects due to fusion α-particles for tokamak reactor design

    International Nuclear Information System (INIS)

    Issues related to the presence of fusion α-particles which are of importance for the design of a tokamak reactor are listed and shortly discussed. It is concluded that these issues, although to a large extent directly connected with the general problems of tokamak physics, require more attention to provide the information needed for designing a tokamak reactor. (orig.)

  2. Development of 3D ferromagnetic model of tokamak core with strong toroidal asymmetry

    DEFF Research Database (Denmark)

    Markovič, Tomáš; Gryaznevich, Mikhail; Ďuran, Ivan;

    2015-01-01

    Fully 3D model of strongly asymmetric tokamak core, based on boundary integral method approach (i.e. characterization of ferromagnet by its surface) is presented. The model is benchmarked on measurements on tokamak GOLEM, as well as compared to 2D axisymmetric core equivalent for this tokamak...

  3. Fast bolometric measurements on the TCV tokamak

    Science.gov (United States)

    Furno, I.; Weisen, H.; Mlynar, J.; Pitts, R. A.; Llobet, X.; Marmillod, Ph.; Pochon, G. P.

    1999-12-01

    The design and first results are presented from a bolometric diagnostic with high temporal resolution recently installed on the TCV tokamak. The system consists of two pinhole cameras viewing the plasma from above and below at the same toroidal location. Each camera is equipped with an AXUV-16ELO linear array of 16 p-n junction photodiodes, characterized by a flat spectral sensitivity from ultraviolet to x-ray energies, a high temporal response (<0.5 μs), and insensitivity to low-energy neutral particles emitted by the plasma. This high temporal resolution allows the study of transient phenomena such as fast magnetohydrodynamic (MHD) activity hitherto inaccessible with standard bolometry. In the case of purely electromagnetic radiation, good agreement has been found when comparing results from the new diagnostic with those from a standard metal foil bolometer system. This comparison has also revealed that the contribution of neutrals to the foil bolometer measurements can be extremely important under certain operating conditions, precluding the application of tomographic techniques for reconstruction of the radiation distribution.

  4. Demonstration tokamak-power-plant study (DEMO)

    International Nuclear Information System (INIS)

    A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li2O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li2O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B4C, lead, and FE 1422 structural material

  5. Aspects of Tokamak toroidal magnet protection

    International Nuclear Information System (INIS)

    Simple but conservative geometric models are used to estimate the potential for damage to a Tokamak reactor inner wall and blanket due to a toroidal magnet field collapse. The ofly potential hazard found to exist is due to the MHD pressure rise in a lithium blanket. A survey is made of proposed protection methods for superconducting torgidal magnets. It is found that the two general classificatigls of protectign methods are thermal and electrical. Computer programs were developed which aldow the toroidal magnet set to be modeled as a set of circular filaments. A simple thermal model of the conductor was used which allows heat transfer to the magnet structure and which includes the effect of temperature dependent properties. To be effective in large magnets an electrical protection system should remove at least 50% of the stored energy in the protection circuit assuming that all of the superconductor in the circuit quenches when the circuit is activated. A protection system design procedure based on this criterion was developed

  6. Diamagnetic measurements on the Alcator C tokamak

    International Nuclear Information System (INIS)

    A procedure for determining the total thermal energy content of a magnetically confined plasma from a measurement of the plasma magnetization has been successfully implemented on the Alcator C tokamak. When a plasma is confined by a magnetic field, the kinetic pressure of the plasma is supported by an interaction between the confining magnetic field and drift currents which flow in the plasma. These drift currents induce an additional magnetic field which can be measured by means of appropriately positioned pickup coils. From a measurement of this magnetic field and of the confining magnetic field, one can calculate the spatially averaged plasma pressure, which is related to the thermal energy content of the plasma by the equation of state of the plasma. The theory on which this measurement is based is described in detail. The fields and currents which flow in the plasma are related to the confining magnetic field and the plasma pressure by requiring that the plasma be in equilibrium, i.e., by balancing the forces due to pressure gradients against those due to magnetic interactions. The apparatus used to make this measurement is described and some example data analyses are carried out

  7. Plasma transport in a Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Nominal predicted plasma conditions in a Compact Ignition Tokamak (CIT) are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models that have given almost equally good fits to experimental data. Using a transport model that best fits the data, thermonuclear ignition occurs in a CIT design with a major radius of 1.32 m, plasma half-width of 0.43 mn, elongation of 2.0, and toroidal field and plasma current ramped in 6 s from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the /sup 3/He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates have a large effect on ignition and on the maximum beta that can be achieved

  8. Physics aspects of the compact ignition tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve nτE ≅ 2 x 1020 s/m3 required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which provides a high level of ohmic heating, improves the energy confinement, and allows a relatively high beta (≅ 6%). The present CIT design also has a high degree of elongation (κ ≅ 1.8) to aid in producing the large plasma current. A double null poloidal divertor and pellet injection are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Auxiliary heating is expected to be necessary to achieve ignition, and 10-20 MW of ICRF is to be provided. (orig.)

  9. Experimental results from the TFTR tokamak

    International Nuclear Information System (INIS)

    Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam-heated discharges. The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2 T). Initial neutral-beam-heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 1020 m-3 in high current discharges. At the opposite extreme, by operating at low plasma current (I/sub p/ ∼ 0.8 MA) and low density (anti n/sub e/ ∼ 1 x 1019 m-3), high ion temperatures (9 +- 2 keV) and rotation speeds (7 x 105 m/s) have been achieved during injection. In addition, plasma compression experiments have demonstrated acceleration of beam ions from 82 keV to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam-heated discharges. The results of these confinement studies are presented

  10. Industry roles in the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) is the first major fusion project opportunity in many years for US industry. Both the TPX management and the Department of Energy's Office of Fusion Energy are committed to creating industry roles that are integrated throughout the project and that appropriately use the capabilities they offer. To address industry roles in TPX it is first appropriate to describe the collaborative national approach taken for this program. The Director of the Princeton Plasma Physics Laboratory (PPPL) was asked by DOE to set up this national team structure, and the current senior management positions and delegated responsibilities reflect that approach. While reporting lines and delegated roles are clear in the organization chart for TPX, one way to view, it, different from that of the individuals responsible upward through this management structure for various elements of the project, is through institutional responsibilities to the senior management team. In this view the management team relies on several national laboratories, each using industry contracts for major sub-systems and components, to execute the project. These responsibilities for design and for contracting are listed, showing that all major contracts will come through three national laboratories, forming teams for their responsible activities

  11. Aspects of Tokamak toroidal magnet protection

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.W.; Kazimi, M.S.

    1979-07-01

    Simple but conservative geometric models are used to estimate the potential for damage to a Tokamak reactor inner wall and blanket due to a toroidal magnet field collapse. The only potential hazard found to exist is due to the MHD pressure rise in a lithium blanket. A survey is made of proposed protection methods for superconducting toroidal magnets. It is found that the two general classifications of protection methods are thermal and electrical. Computer programs were developed which allow the toroidal magnet set to be modeled as a set of circular filaments. A simple thermal model of the conductor was used which allows heat transfer to the magnet structure and which includes the effect of temperature dependent properties. To be effective in large magnets an electrical protection system should remove at least 50% of the stored energy in the protection circuit assuming that all of the superconductor in the circuit quenches when the circuit is activated. A protection system design procedure based on this criterion was developed.

  12. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m2 and a particle heat flux of 1 MW/m2. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  13. Empirical particle transport model for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Petravic, M.; Kuo-Petravic, G.

    1986-08-01

    A simple empirical particle transport model has been constructed with the purpose of gaining insight into the L- to H-mode transition in tokamaks. The aim was to construct the simplest possible model which would reproduce the measured density profiles in the L-regime, and also produce a qualitatively correct transition to the H-regime without having to assume a completely different transport mode for the bulk of the plasma. Rather than using completely ad hoc constructions for the particle diffusion coefficient, we assume D = 1/5 chi/sub total/, where chi/sub total/ approx. = chi/sub e/ is the thermal diffusivity, and then use the kappa/sub e/ = n/sub e/chi/sub e/ values derived from experiments. The observed temperature profiles are then automatically reproduced, but nontrivially, the correct density profiles are also obtained, for realistic fueling rates and profiles. Our conclusion is that it is sufficient to reduce the transport coefficients within a few centimeters of the surface to produce the H-mode behavior. An additional simple assumption, concerning the particle mean-free path, leads to a convective transport term which reverses sign a few centimeters inside the surface, as required by the H-mode density profiles.

  14. Physics aspects of the Compact Ignition Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Post, D.; Bateman, G.; Houlberg, W.; Bromberg, L.; Cohn, D.; Colestock, P.; Hughes, M.; Ignat, D.; Izzo, R.; Jardin, S.

    1986-11-01

    The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha-particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve the condition of ntau/sub E/ approx. 2 x 10/sup 20/ sec m/sup -3/ required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which improves the energy confinement, and provides a high level of ohmic heating. The present CIT design also has a gigh degree of elongation (k approx. 1.8) to aid in producing the large plasma current. A double null poloidal divertor and a pellet injector are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Since auxiliary heating is expected to be necessary to achieve ignition, 10 to 20 MW of Ion Cyclotron Radio Frequency (ICRF) is to be provided.

  15. Neoclassical transport of impurtities in tokamak plasmas

    International Nuclear Information System (INIS)

    Tokamak plasmas are inherently comprised of multiple ion species. This is due to wall-bred impurities and, in future reactors, will result from fusion-born alpha particles. Relatively small concentrations of highly charged non-hydrogenic impurities can strongly influence plasma transport properties whenever n/sub I/e/sub I/2/n/sub H/e2 greater than or equal to (m/sub e//m/sub H/)/sup 1/2/. The determination of the complete neoclassical Onsager matrix for a toroidally confined multispecies plasma, which provides the linear relation between the surface averaged radial fluxes and the thermodynamic forces (i.e., gradients of density and temperature, and the parallel electric field), is reviewed. A closed set of one-dimensional moment equations is presented for the time evolution of thermodynamic and magnetic field quantities which results from collisional transport of the plasma and two dimensional motion of the magnetic flux surface geometry. The effects of neutral beam injection on the equilibrium and transport properties of a toroidal plasma are consistently included

  16. Control of the vertical instability in tokamaks

    International Nuclear Information System (INIS)

    The problem of control of the vertical instability is formulated for a massless filamentary plasma. The massless approximation is justified by an examination of the role of inertia in the control problem. The system is solved using Laplace transform techniques. The linear system is studied to determine the stability boundaries. It is found that the system can be stabilized up to a critical decay index, which is predominantly a function of the geometry of the passive stabilizing shell. A second, smaller critical index, which is a function of the geometry of the control coils, determines the limit of stability in the absence of derivative gain in the control circuit. The system is also studied numerically in order to incorporate the non-linear effects of power supply dynamics. The power supply bandwidth requirement is determined by the open-loop growth rate of the instability. The system is studied for a number of control coil options which are available on the DIII-D tokamak. It is found that many of the coils will not provide adequate stabilization and that the use of inboard coils is advantageous in stabilizing the system up to the critical index. Experiments carried out on DIII-D confirm the appropriateness of the model. Using the results of the model study, we have stabilized DIII-D plasmas with decay indices up to 98% of the critical index. Measurement of the plasma vertical position is also discussed. (author) 27 figs., 6 refs

  17. Neoclassical transport in high {beta} tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high {beta} large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high {beta} large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low {beta} values by a factor ({var_epsilon}/q{sup 2}{beta}){sup {1/2}} II. This factor is the ratio of plasma volume in the boundary layer to the volume in the core. The fraction of trapped particles on a given flux surface (f{sub t}) is also reduced by this factor so that {approximately} {sub ({var_epsilon}}/q{sup 2}{beta}){sup {1/2}}. Special attention is given to the current equation, since this is thought to be relevant at low 3 and therefore may also be relevant at high {beta}. The bootstrap current term is found to exceed the actual current by a factor of the square root of the aspect ratio.

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

    Science.gov (United States)

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

    2011-10-01

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

  19. TSC (Tokamak Simulation Code) disruption scenarios and CIT (Compact Ignition Tokamak) vacuum vessel force evolution

    Energy Technology Data Exchange (ETDEWEB)

    Sayer, R.O.; Peng, Y.K.M.; Strickler, D.J.; Jardin, S.C.

    1990-01-01

    The Tokamak Simulation Code and the TWIR postprocessor code have been used to develop credible plasma disruption scenarios for the Compact Ignition Tokamak (CIT) in order to predict the evolution of forces on CIT conducting structures and to provide results required for detailed structural design analysis. The extreme values of net radial and vertical vacuum vessel (VV) forces were found to be F{sub R}={minus}12.0 MN/rad and F{sub Z}={minus}3.0 MN/rad, respectively, for the CIT 2.1-m, 11-MA design. Net VV force evolution was found to be altered significantly by two mechanisms not noted previously. The first, due to poloidal VV currents arising from increased plasma paramagnetism during thermal quench, reduces the magnitude of the extreme F{sub R} by 15-50{percent} and modifies the distribution of forces substantially. The second effect is that slower plasma current decay rates give more severe net vertical VV loads because the current decay occurs when the plasma has moved farther from midplane than is the case for faster decay rates. 7 refs., 9 figs., 1 tab.

  20. TSC [Tokamak Simulation Code] disruption scenarios and CIT [Compact Ignition Tokamak] vacuum vessel force evolution

    International Nuclear Information System (INIS)

    The Tokamak Simulation Code and the TWIR postprocessor code have been used to develop credible plasma disruption scenarios for the Compact Ignition Tokamak (CIT) in order to predict the evolution of forces on CIT conducting structures and to provide results required for detailed structural design analysis. The extreme values of net radial and vertical vacuum vessel (VV) forces were found to be FR=-12.0 MN/rad and FZ=-3.0 MN/rad, respectively, for the CIT 2.1-m, 11-MA design. Net VV force evolution was found to be altered significantly by two mechanisms not noted previously. The first, due to poloidal VV currents arising from increased plasma paramagnetism during thermal quench, reduces the magnitude of the extreme FR by 15-50% and modifies the distribution of forces substantially. The second effect is that slower plasma current decay rates give more severe net vertical VV loads because the current decay occurs when the plasma has moved farther from midplane than is the case for faster decay rates. 7 refs., 9 figs., 1 tab

  1. Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.

  2. Deposit of thin films for Tokamaks conditioning; Deposito de peliculas delgadas para acondicionar Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Valencia A, R

    2006-07-01

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (<10 eV), both in noble and reactive gases, as well as the conditioning by thin film deposits of hydrogen rich amorphous carbon (carbonization) leading to a reduction in the plasma resistivity from 8.99 x 10{sup -6} to 4.5 x 10{sup -6} {omega}-m, thus taking the Z{sub ef} value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission

  3. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

    Boncagni, Luca, E-mail: luca.boncagni@enea.it [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Pucci, Daniele; Piesco, F.; Zarfati, Emanuele [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy); Mazzitelli, G. [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Monaco, S. [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy)

    2013-10-15

    Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1].

  4. Hybrid Method for Tokamak MHD Equilibrium Configuration Reconstruction

    Institute of Scientific and Technical Information of China (English)

    HE Hong-Da; DONG Jia-Qi; ZHANG Jin-Hua; JIANG Hai-Bin

    2007-01-01

    A hybrid method for tokamak MHD equilibrium configuration reconstruction is proposed and employed in the modified EFIT code. This method uses the free boundary tokamak equilibrium configuration reconstruction algorithm with one boundary point fixed. The results show that the position of the fixed point has explicit effects on the reconstructed divertor configurations. In particular, the separatrix of the reconstructed divertor configuration precisely passes the required position when the hybrid method is used in the reconstruction. The profiles of plasma parameters such as pressure and safety factor for reconstructed HL-2A tokamak configurations with the hybrid and the free boundary methods are compared. The possibility for applications of the method to swing the separatrix strike point on the divertor target plate is discussed.

  5. The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

    Directory of Open Access Journals (Sweden)

    Jing Li

    2014-01-01

    Full Text Available The stability and bifurcations of multiple limit cycles for the physical model of thermonuclear reaction in Tokamak are investigated in this paper. The one-dimensional Ginzburg-Landau type perturbed diffusion equations for the density of the plasma and the radial electric field near the plasma edge in Tokamak are established. First, the equations are transformed to the average equations with the method of multiple scales and the average equations turn to be a Z2-symmetric perturbed polynomial Hamiltonian system of degree 5. Then, with the bifurcations theory and method of detection function, the qualitative behavior of the unperturbed system and the number of the limit cycles of the perturbed system for certain groups of parameter are analyzed. At last, the stability of the limit cycles is studied and the physical meaning of Tokamak equations under these parameter groups is given.

  6. Fractal structure of films deposited in a tokamak

    Science.gov (United States)

    Budaev, V. P.; Khimchenko, L. N.

    2007-04-01

    The surface of amorphous films deposited in the T-10 tokamak was studied in a scanning tunnel microscope. The surface relief on a scale from 10 nm to 100 μm showed a stochastic surface topography and revealed a hierarchy of grains. The observed variety of irregular structures of the films was studied within the framework of the concept of scale invariance using the methods of fractal geometry and statistical physics. The experimental probability density distribution functions of the surface height variations are close in shape to the Cauchy distribution. The stochastic surface topography of the films is characterized by a Hurst parameter of H = 0.68-0.85, which is evidence of a nontrivial self-similarity of the film structure. The fractal character and porous structure of deposited irregular films must be considered as an important issue related to the accumulation of tritium in the ITER project. The process of film growth on the surface of tokamak components exposed to plasma has been treated within the framework of the general concept of inhomogeneous surface growth. A strong turbulence of the edge plasma in tokamaks can give rise to fluctuations in the incident flux of particles, which leads to the growth of fractal films with grain dimensions ranging from nano-to micrometer scale. The shape of the surface of some films found in the T-10 tokamak has been interpreted using a model of diffusion-limited aggregation (DLA). The growth of films according to the discrete DLA model was simulated using statistics of fluctuations observed in a turbulent edge plasma of the T-10 tokamak. The modified DLA model reproduces well the main features of the surface of some films deposited in tokamaks.

  7. Simulation of EAST vertical displacement events by tokamak simulation code

    Science.gov (United States)

    Qiu, Qinglai; Xiao, Bingjia; Guo, Yong; Liu, Lei; Xing, Zhe; Humphreys, D. A.

    2016-10-01

    Vertical instability is a potentially serious hazard for elongated plasma. In this paper, the tokamak simulation code (TSC) is used to simulate vertical displacement events (VDE) on the experimental advanced superconducting tokamak (EAST). Key parameters from simulations, including plasma current, plasma shape and position, flux contours and magnetic measurements match experimental data well. The growth rates simulated by TSC are in good agreement with TokSys results. In addition to modeling the free drift, an EAST fast vertical control model enables TSC to simulate the course of VDE recovery. The trajectories of the plasma current center and control currents on internal coils (IC) fit experimental data well.

  8. Injection of intense ion beam into a tokamak

    International Nuclear Information System (INIS)

    We describe an experiment to investigate the direct injection of an intense ion beam into a tokamak by means of the polarization drift. Confinement of 100 keV ions in the UCI tokamak (r = 15 cm, R = 60 cm, B/sub T/ = 6 kG) requires operation with a plasma current of 56 kA corresponding to q (limiter) = 2. Trapped ions are to be detected by a charge-exchange analyzer. The present status of the experiment will be discussed

  9. Soft X-ray tomography on HT-6B tokamak

    International Nuclear Information System (INIS)

    The tomography method for deriving soft X-ray local emissivities on HT-6B tokamak, using one horizontal array of 23 soft X-ray detectors, is described. This method has been applied to study of sawtooth oscillation and large m = 1 oscillation on HT-6B tokamak. It has been found that the large m = 1 oscillation on soft X-ray signal is caused by the rotation of plasma column containing perturbation. The reconstructed images in the phase before a sawtooth crash have shown the hot plasma core move gradually toward one side, this is considered that as the increasing m=1 kink mode. (author). 5 refs, 7 figs

  10. Fusion-product transport in axisymmetric tokamaks: losses and thermalization

    International Nuclear Information System (INIS)

    High-energy fusion-product losses from an axisymmetric tokamak plasma are studied. Prompt-escape loss fluxes (i.e. prior to slowing down) are calculated including the non-separable dependence of flux as a function of poloidal angle and local angle-of-incidence at the first wall. Fusion-product (fp) thermalization and heating are calculated assuming classical slowing down. The present analytical model describes fast ion orbits and their distribution function in realistic, high-β, non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated

  11. Data processing system for spectroscopy at Novillo Tokamak

    International Nuclear Information System (INIS)

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  12. Stability of Tokamaks with respect to slip motions

    International Nuclear Information System (INIS)

    Using the energy principle in Tokamaks we investigate a class of perturbations which, if unstable, cannot be stabilized by the toroidal main field. On the assumptions of usual Tokamak ordering and in the limit of infinite aspect ratio, these perturbations are shown to be minimizing among all axisymmetric perturbations. In the case of finite aspect ratio, a detailed stability analysis is carried out using a constant pressure surface current model with elliptic, triangular or rectangular plasma cross-section. Definite stabilization by toroidal effects and by beta poloidal is demonstrated. (orig.)

  13. A CONCEPT FOR NEXT STEP ADVANCED TOKAMAK FUSION DEVICE

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    A concept is introduced for initiating the design study of a special class of tokamak,which has a magnetic confinement configuration intermediate between contemporary advanced tokamak and the recently established spherical torus (ST,also well known by the name "spherical tokamak").The leading design parameter in the present proposal is a dimensionless geometrical parameter, the machine aspect ratio A=R0/a0=2.0,where the parameters a0 and R0 denote,respectively,the plasma (equatorial) minor radius and the plasma major radius.The aim of this choice is to technologically and experimentally go beyond the aspect ratio frontier (R0/a0≈2.5) of present day tokamaks and enter a broad unexplored domain existing on the (a0,R0) parameter space in current international tokamak database,between the data region already moderately well covered by the advanced conventional tokamaks and the data region planned to be covered by STs.Plasma minor radius a0 has been chosen to be the second basic design parameter, and consequently,the plasma major radius R0 is regarded as a dependent design parameter.In the present concept,a nominal plasma minor radius a0=1.2m is adopted to be the principal design value,and smaller values of a0 can be used for auxiliary design purposes,to establish extensive database linkage with existing tokamaks.Plasma minor radius can also be adjusted by mechanical and/or electromagnetic means to smaller values during experiments,for making suitable data linkages to existing machines with higher aspect ratios and smaller plasma minor radii.The basic design parameters proposed enable the adaptation of several confinement techniques recently developed by STs,and thereby a specially arranged central-bore region inside the envisioned tokamak torus,with retrieved space in the direction of plasma minor radius,will be available for technological adjustments and maneuverings to facilitate implementation of engineering instrumentation and real time high

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

    International Nuclear Information System (INIS)

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

  15. Adaptive grid finite element model of the tokamak scrapeoff layer

    Energy Technology Data Exchange (ETDEWEB)

    Kuprat, A.P.; Glasser, A.H. [Los Alamos National Lab., NM (United States)

    1995-07-01

    The authors discuss unstructured grids for application to transport in the tokamak edge SOL. They have developed a new metric with which to judge element elongation and resolution requirements. Using this method, the authors apply a standard moving finite element technique to advance the SOL equations while inserting/deleting dynamically nodes that violate an elongation criterion. In a tokamak plasma, this method achieves a more uniform accuracy, and results in highly stretched triangular finite elements, except near separatrix X-point where transport is more isotropic.

  16. What is the fate of runaway positrons in tokamaks?

    International Nuclear Information System (INIS)

    Massive runaway positrons are generated by runaway electrons in tokamaks. The fate of these positrons encodes valuable information about the runaway dynamics. The phase space dynamics of a runaway position is investigated using a Lagrangian that incorporates the tokamak geometry, loop voltage, radiation and collisional effects. It is found numerically that runaway positrons will drift out of the plasma to annihilate on the first wall, with an in-plasma annihilation possibility less than 0.1%. The dynamics of runaway positrons provides signatures that can be observed as diagnostic tools

  17. Multichannel bolometer for radiation measurements on the TCA tokamak

    International Nuclear Information System (INIS)

    A multichannel radiation bolometer has been developed for the Tokamak Chauffage Alfven (TCA) tokamak. It has 16 equally spaced chords that view the plasma through a narrow horizontal slit. Almost an entire vertical plasma cross section can be observed. The bolometer operates on the basis of a semiconducting element which serves as a temperature-dependent resistance. A new electronic circuit has been developed which takes advantage of the semiconductor characteristics of the detector by using feedback techniques. Measurements made with this instrument are discussed

  18. Design and Analysis of the Thermal Shield of EAST Tokamak

    Science.gov (United States)

    Xie, Han; Liao, Ziying

    2008-04-01

    EAST (Experimental Advanced Superconducting Tokamak) is a tokamak with superconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.

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

    International Nuclear Information System (INIS)

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

  20. Atomic data for integrated tokamak modelling

    International Nuclear Information System (INIS)

    The Integrated Tokamak Modeling Task Force (ITM-TF) was set up in 2004. The main target is to coordinate the European fusion modeling effort and providing a complete European modeling structure for International Thermonuclear Experimental Reactor (ITER), with the highest degree of flexibility. For the accurate simulation of the processes in the active fusion reactor in the ITM-TF, numerous atomic, molecular, nuclear and surface related data are required. In this work we present total-, single- and multiple-ionization and charge exchange cross sections in close connection to the ITM-TF. Interpretation of these cross sections in multi-electron ion-atom collisions is a challenging task for theories. The main difficulty is caused by the many-body feature of the collision, involving the projectile, projectile electron(s), target nucleus, and target electron(s). The classical trajectory Monte Carlo (CTMC) method has been quite successful in dealing with the atomic processes in ion-atom collisions. One of the advantages of the CTMC method is that many-body interactions are exactly taken into account related CTMC simulations for a various collision systems are presented. To highlight the efficiency of the method we present electron emission cross sections in collision between dressed Alq+ ions with He target. The theory delivers separate spectra for electrons emitted from the target and the projectile. By summing these two components in the rest frame of the target we may make a comparison with available experimental data. For the collision system in question, a significant contribution from Fermi-shuttle ionization has to be expected in the spectra at energies higher than E=0.5 me (nV)2, where me is the mass of the electron, V the projectile velocity and n an integer greater than 1. We found enhanced electron yields compared to first order theory in this region of CTMC spectra, which can be directly attributed to the contribution of Fermi-shuttle type multiple scattering

  1. Magnet systems for ''Bean-Shaped'' tokamak

    International Nuclear Information System (INIS)

    Bean-shaping of tokamak plasmas offers a method of reaching stable operation at (beta) > 10%. In order to establish the indentation of the ''bean'', a set of high- current ''pushing coils'' (> 5 MA in a reactor) must be located at the midplane as close as possible to the inboard edge of the plasma. If located in the bore of the TF coils, then maintenance of the pushing coils may be impossible, and the interlocking coils may prevent reactor modularity. If located outside, the required pushing-coil current may be unacceptably large. This dilemma is overcome with a unique TF coil design in which the inboard leg is bent outward in the form of an arc. The pushing coils are housed in the midplane indentation of this arc, just outside the TF coils but adequately close to the plasma. The arched coil transfers forces to the top and bottom legs, where it can be reacted by a clamp structure if necessary. This technique would allow demountable joints to be placed near the inoard leg (for copper TF coils). Another design approach to the pushing coils is to use liquid Li or Na as the conductor and coolant. The liquid metal ''coils'' can be placed immediately adjacent to the plasma, giving optimal control of the plasma shape with minimal coil current, although modularity of the reactor may have to be surrendered. Conceptual designs are presented of PF and TF coil systems for an ignition test reactor with about 14% and for a full-scale demonstration reactor with about 20%, both using copper TF coils

  2. Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  3. Ion cyclotron emission in tokamak plasmas; Emission cyclotronique ionique dans les plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Fraboulet, D.

    1996-09-17

    Detection of {alpha}(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, {alpha} particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. {alpha} particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central {alpha} density in a reactor. (author). 166 refs.

  4. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author)

  5. Modelling multi-ion plasma gun simulations of Tokamak disruptions

    International Nuclear Information System (INIS)

    The effect of impurity ions in plasma gun ablation tests of various targets is considered. Inclusion of reasonable amounts of impurity (∼10%) is adequate to explain observed energy transmission and erosion measurements. The gun tests and the computer code calculations are relevant to the parameter range expected for major disruptions on large tokamaks

  6. Economic evaluation of fissile fuel production using resistive magnet tokamaks

    International Nuclear Information System (INIS)

    The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U3O8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U3O8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

  7. Conceptual design of Remote Control System for EAST tokamak

    International Nuclear Information System (INIS)

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

  8. Evidence of Inward Toroidal Momentum Convection in the JET Tokamak

    DEFF Research Database (Denmark)

    Tala, T.; Zastrow, K.-D.; Ferreira, J.;

    2009-01-01

    Experiments have been carried out on the Joint European Torus tokamak to determine the diffusive and convective momentum transport. Torque, injected by neutral beams, was modulated to create a periodic perturbation in the toroidal rotation velocity. Novel transport analysis shows the magnitude an...

  9. Bulk Ion Heating with ICRF Waves in Tokamaks

    DEFF Research Database (Denmark)

    Mantsinen, M. J.; Bilato, R.; Bobkov, V. V.;

    2015-01-01

    Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER a...

  10. INTOR: a first-generation tokamak experimental reactor

    International Nuclear Information System (INIS)

    An intensive, year-long, international evaluation of the next major tokamak beyond the generation of large experiments currently under construction was carried out during 1979. This evaluation consisted of the definition of objectives, an assessment of the physics and technology base and R and D needs and the identification of a set of parameters that physically characterize the machine

  11. Evaluation of the average ion approximation for a tokamak plasma

    International Nuclear Information System (INIS)

    The average ion approximation, sometimes used to calculated atomic processes in plasmas, is assessed by computing deviations in various rates over a set of conditions representative of tokamak edge plasmas. Conditions are identified under which the rates are primarily a function of the average ion charge and plasma parameters, as assumed in the average ion approximation. (Author) 19 refs., tab., 5 figs

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

    International Nuclear Information System (INIS)

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

  13. Gas breakers for tokamak OHMIC-heating duty

    International Nuclear Information System (INIS)

    The current interrupting capacity of air blast and SF6 breakers is reviewed for application in tokamak ohmic-heating circuits. Particular attention is paid to generator breakers for their large current interrupting capacity and suitability for ohmic-heating circuits

  14. Feedback Control for Plasma Position on HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    LIBo; SONGXianming; LILi; LIULi; WANGMinghong; FANMingjie; CHENLiaoyuan; YAOLieying; YANGQingwei

    2003-01-01

    HL-2A is a tokamak with closed divertor. It had been built at the end of 2002 and began to discharge from then on. To further study plasma discharges in HL-2A, a feedback control system (FBCS) for plasma position bad been developed in 2003.

  15. Performance and development of the DIII-D tokamak core

    International Nuclear Information System (INIS)

    The DIII-D tokamak is an upgrade of the Doublet III configuration which has operated since early 1986. This paper presents recent advances in performance using the upper divertor, fabrication development for vanadium components, operation of the helium leak checking in a high deuterium background, and restoration of the damaged Ohmic heating solenoid

  16. A synchronization system to digitalize TJ-1 Tokamak data

    International Nuclear Information System (INIS)

    At TJ-1 Tokamak signals are stored on a 60-channel magnetic memory. In this report, a system to address those channels and synchronize readout is presented. Digitalized signals are stored in structured files on PDP-11/34 magnetic disks. (author)

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

    Science.gov (United States)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

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

  18. Investigation of the microwave emission from the PRETEXT tokamak

    International Nuclear Information System (INIS)

    A study of the microwave emission from the PRETEXT tokamak has been conducted. Two types of emission have been observed: electron cyclotron and electron plasma frequency. Three general emission regimes have been identified. These regimes are best classified by the dimensionless parameter α, where α = ω/sub pe//Ω/sub e/

  19. Gamma ray imager on the DIII-D tokamak.

    Science.gov (United States)

    Pace, D C; Cooper, C M; Taussig, D; Eidietis, N W; Hollmann, E M; Riso, V; Van Zeeland, M A; Watkins, M

    2016-04-01

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1-60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons. PMID:27131674

  20. Conceptual design of Remote Control System for EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  1. Dynamic diagnostics of the error fields in tokamaks

    Science.gov (United States)

    Pustovitov, V. D.

    2007-07-01

    The error field diagnostics based on magnetic measurements outside the plasma is discussed. The analysed methods rely on measuring the plasma dynamic response to the finite-amplitude external magnetic perturbations, which are the error fields and the pre-programmed probing pulses. Such pulses can be created by the coils designed for static error field correction and for stabilization of the resistive wall modes, the technique developed and applied in several tokamaks, including DIII-D and JET. Here analysis is based on the theory predictions for the resonant field amplification (RFA). To achieve the desired level of the error field correction in tokamaks, the diagnostics must be sensitive to signals of several Gauss. Therefore, part of the measurements should be performed near the plasma stability boundary, where the RFA effect is stronger. While the proximity to the marginal stability is important, the absolute values of plasma parameters are not. This means that the necessary measurements can be done in the diagnostic discharges with parameters below the nominal operating regimes, with the stability boundary intentionally lowered. The estimates for ITER are presented. The discussed diagnostics can be tested in dedicated experiments in existing tokamaks. The diagnostics can be considered as an extension of the 'active MHD spectroscopy' used recently in the DIII-D tokamak and the EXTRAP T2R reversed field pinch.

  2. Spatially resolved soft x-ray spectroscopy of tokamak plasmas

    International Nuclear Information System (INIS)

    We describe the space-resolved soft x-ray (1-33nm) instrumentation developed for the Tore Supra tokamak. By using a programmable hydraulic jack to move the spectrometer, several spatial profiles (up to ten) of many impurity lines are obtained during a single plasma discharge, with a time resolution which can be as short as 600 ms. (author)

  3. Tokamak Scenario Trajectory Optimization Using Fast Integrated Simulations

    Science.gov (United States)

    Urban, Jakub; Artaud, Jean-François; Vahala, Linda; Vahala, George

    2015-11-01

    We employ a fast integrated tokamak simulator, METIS, for optimizing tokamak discharge trajectories. METIS is based on scaling laws and simplified transport equations, validated on existing experiments and capable of simulating a full tokamak discharge in about 1 minute. Rapid free-boundary equilibrium post-processing using FREEBIE provides estimates of PF coil currents or forces. We employ several optimization strategies for optimizing key trajectories, such as Ip or heating power, of a model ITER hybrid discharge. Local and global algorithms with single or multiple objective functions show how to reach optimum performance, stationarity or minimum flux consumption. We constrain fundamental operation parameters, such as ramp-up rate, PF coils currents and forces or heating power. As an example, we demonstrate the benefit of current over-shoot for hybrid mode, consistent with previous results. This particular optimization took less than 2 hours on a single PC. Overall, we have established a powerful approach for rapid, non-linear tokamak scenario optimization, including operational constraints, pertinent to existing and future devices design and operation.

  4. The Effect of Recycling in the HL-1M Tokamak

    Institute of Scientific and Technical Information of China (English)

    ZHENGYongzhen

    2002-01-01

    It is often stated that even clean tokamak discharges disrupt at high density. One possibility is that such disruption result from the energy loss arising from hydrogen recycling at the edge of the plasma.this energy loss could lead to a contraction of the current channel and the production of a disruptively unstable configuration.

  5. Stability of infernal and ballooning modes in advanced tokamak scenarios

    NARCIS (Netherlands)

    Holties, H. A.; Huysmans, G. T. A.; Goedbloed, J. P.; Kerner, W.; Parail, V.V.; Soldner, F. X.

    1996-01-01

    A numerical parameter study has been performed in order to find MHD stable operating regimes for advanced tokamak experiments In this study we have concentrated on internal modes. Ballooning stability and stability with respect to infernal modes are considered. The calculations confirm that pressure

  6. General Description of Ideal Tokamak MHD Instability Ⅱ

    Institute of Scientific and Technical Information of China (English)

    石秉仁

    2002-01-01

    In this subsequent study on general description of ideal tokamak MHD instability,the part Ⅱ, by using a coordinate with rectified magnetic field lines, the eigenmode equationsdescribing the low-mode-number toroidal Alfven modes (TAE and EAE) are derived through afurther expansion of the shear Alfven equation of motion.

  7. Fokker-Planck Study of Tokamak Electron Cyclotron Resonance Heating

    Institute of Scientific and Technical Information of China (English)

    SHIBingren; LONGYongxing; DONGJiaqi; LIWenzhong; JIAOYiming; WANGAike

    2002-01-01

    In this study, we add a subroutine for describing the electron cyclotron resonant heating calculation to the Fokker-Planck code. By analyzing the wave-particle resonance condition in tokamak plasma and the fast motion of electrons along magnetic field lines, suitable quasi-linear diffusion coefficients are given.

  8. MHD analysis of edge instabilities in the JET tokamak

    NARCIS (Netherlands)

    Perez von Thun, Christian Pedro

    2004-01-01

    The aim of nuclear fusion energy research is to demonstrate the feasibility of nuclear fusion reactors as a future energy source. The tokamak is the most advanced fusion machine to date, and is most likely the first system to be converted into a reactor. An important subject of nuclear fusion resear

  9. Disruption avoidance through active magnetic feedback in tokamak plasmas

    Science.gov (United States)

    Paccagnella, Roberto; Zanca, Paolo; Yanovskiy, Vadim; Finotti, Claudio; Manduchi, Gabriele; Piron, Chiara; Carraro, Lorella; Franz, Paolo; RFX Team

    2014-10-01

    Disruptions avoidance and mitigation is a fundamental need for a fusion relevant tokamak. In this paper a new experimental approach for disruption avoidance using active magnetic feedback is presented. This scheme has been implemented and tested on the RFX-mod device operating as a circular tokamak. RFX-mod has a very complete system designed for active mode control that has been proved successful for the stabilization of the Resistive Wall Modes (RWMs). In particular the current driven 2/1 mode, unstable when the edge safety factor, qa, is around (or even less than) 2, has been shown to be fully and robustly stabilized. However, at values of qa (qa > 3), the control of the tearing 2/1 mode has been proved difficult. These results suggested the idea to prevent disruptions by suddenly lowering qa to values around 2 where the tearing 2/1 is converted to a RWM. Contrary to the universally accepted idea that the tokamaks should disrupt at low qa, we demonstrate that in presence of a well designed active control system, tokamak plasmas can be driven to low qa actively stabilized states avoiding plasma disruption with practically no loss of the plasma internal energy.

  10. High- Q plasmas in the TFTR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.; Barnes, C.W.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Dylla, H.F.; Efthimion, P.C.; Fredrickson, E.D.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.C.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S.J.; LaMarche, P.H.; LeBlanc, B.; Mansfield, D.K.; Marmar, E.S.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.K.; Paul, S.F.; Pitcher, S.; Ramsey, A.T.; Redi, M.H.; Sabbagh, S.A.; Scott, S.D.; Snipes, J.; Stevens, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Terry, J.L.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J. (Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (USA))

    1991-08-01

    In the Tokamak Fusion Test Reactor (TFTR) (Plasma Phys. Controlled Fusion {bold 26}, 11 (1984)), the highest neutron source strength {ital S}{sub {ital n}} and D--D fusion power gain {ital Q}{sub DD} are realized in the neutral-beam-fueled and heated supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, {ital S}{sub {ital n}} increases approximately as {ital P}{sup 1.8}{sub {ital b}}. The highest-{ital Q} shots are characterized by high {ital T}{sub {ital e}} (up to 12 keV), {ital T}{sub {ital i}} (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad {ital T}{sub {ital e}} profiles, and lower {ital Z}{sub eff}. Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, {ital Q}{sub DD} increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness ({ital n}{sub {ital e}}(0)/{l angle}{ital n}{sub {ital e}}{r angle}) during the beam pulse. To date, the best fusion results are {ital S}{sub {ital n}}=5{times}10{sup 16} n/sec, {ital Q}{sub DD}=1.85{times}10{sup {minus}3}, and neutron yield=4.0{times}10{sup 16} n/pulse, obtained at {ital I}{sub {ital p}}=1.6--1.9 MA and beam energy {ital E}{sub {ital b}}=95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of {ital S}{sub {ital n}} arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with {ital P}{sub {ital b}}.

  11. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

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

  12. Design and construction of Alborz tokamak vacuum vessel system

    Energy Technology Data Exchange (ETDEWEB)

    Mardani, M., E-mail: mohsenmardani@gmail.com [Amirkabir University of Technology (Tehran Polytechnic), Tehran (Iran, Islamic Republic of); Amrollahi, R.; Koohestani, S. [Amirkabir University of Technology (Tehran Polytechnic), Tehran (Iran, Islamic Republic of)

    2012-09-15

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

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

    International Nuclear Information System (INIS)

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

  14. Protection of tokamak plasma facing components by a capillary porous system with lithium

    Science.gov (United States)

    Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

    2015-08-01

    Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

  15. Fueling studies on the lithium tokamak experiment

    Science.gov (United States)

    Lundberg, Daniel Patrick

    Lithium plasma facing components reduce the flux of "recycled" particles entering the plasma edge from the plasma facing components. This results in increased external fueling requirements and provides the opportunity to control the magnitude and distribution of the incoming particle flux. It has been predicted that the plasma density profile will then be determined by the deposition profile of the external fueling, rather than dominated by the recycled particle flux. A series of experiments on the Lithium Tokamak Experiment demonstrate that lithium wall coatings facilitate control of the neutral and plasma particle inventories. With fresh lithium coatings and careful gas injection programming, over 90% of the injected particle inventory can be absorbed in the lithium wall during a discharge. Furthermore, dramatic changes in the fueling requirements and plasma parameters were observed when lithium coatings were applied. This is largely due to the elimination of water as an impurity on the plasma facing components. A Molecular Cluster Injector (MCI) was developed for the fueling of LTX plasmas. The MCI uses a supersonic nozzle, cooled to liquid nitrogen temperatures, to create the conditions necessary for molecular cluster formation. It has been predicted that molecular clusters will penetrate deeper into plasmas than gas-phase molecules via a reduced ionization cross-section and by improving the collimation of the neutral jet. Using an electron beam diagnostic, the densities of the cryogenic MCI are measured to be an order of magnitude higher than in the room-temperature jets formed with the same valve pressure. This indicates increased collimation relative to what would be expected from ideal gas dynamics alone. A systematic study of the fueling efficiencies achieved with the LTX fueling systems is presented. The fueling efficiency of the Supersonic Gas Injector (SGI) is demonstrated to be strongly dependent on the distance between the nozzle and plasma edge. The

  16. Turbulent ion heating in TCV Tokamak plasmas

    International Nuclear Information System (INIS)

    The Tokamak à configuration variable (TCV) features the highest electron cyclotron wave power density available to resonantly heat (ECRH) the electrons and to drive noninductive currents in a fusion grade plasma (ECCD). In more than 15 years of exploitation, much effort has been expended on real and velocity space engineering of the plasma electron energy distribution function and thus making electron physics a major research contribution of TCV. When a plasma was first subjected to ECCD, a surprising energisation of the ions, perpendicular to the confining magnetic field, was observed on the charge exchange spectrum measured with the vertical neutral particle analyser (VNPA). It was soon concluded that the ion acceleration was not due to power equipartition between electrons and ions, which, due to the absence of direct ion heating on TCV, has thus far been considered as the only mechanism heating the ions. However, although observed for more than ten years, little attention was paid to this phenomenon, whose cause has remained unexplained to date. The key subject of this thesis is the experimental study of this anomalous ion acceleration, the characterisation in terms of relevant parameters and the presentation of a model simulation of the potential process responsible for the appearance of fast ions. The installation of a new compact neutral particle analyser (CNPA) with an extended high energy range (≥ 50 keV) greatly improved the fast ion properties diagnosis. The CNPA was commissioned and the information derived from its measurement (ion temperature and density, isotopic plasma composition) was validated against other ion diagnostics, namely the active carbon charge exchange recombination spectroscopy system (CXRS) and a neutron counter. In ohmic plasmas, where the ion heating agrees with classical theory, the radial ion temperature profile was successfully reconstructed by vertically displacing the plasma across the horizontal CNPA line of sight. Active

  17. Plasma discharge in ferritic first wall vacuum vessel of the Hitachi Tokamak HT-2

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio [Hitachi Ltd., Tokyo (Japan)

    1997-11-01

    A tokamak discharge with ferritic material first wall was tried successfully. The Hitachi Tokamak HT-2 had a stainless steel SUS304 vacuum vessel and modified to have a ferritic plate first wall for experiments to investigate the possibility of ferritic material usage in magnetic fusion devices. The achieved vacuum pressure and times used for discharge cleaning was roughly identical with the stainless steel first wall or the original HT-2. We concluded that ferritic material vacuum vessel is possible for tokamaks. (author)

  18. Configuration studies for a small-aspect-ratio tokamak stellarator hybrid

    International Nuclear Information System (INIS)

    The use of modulated toroidal coils offers a new path to the tokamak-stellarator hybrids. Low-aspect-ratio configurations can be found with robust vacuum flux surfaces and rotational transform close to the transform of a reverse-shear tokamak. These configurations have clear advantages in minimizing disruptions and their effect and in reducing tokamak current drive needs. They also allow the study of low-aspect-ratio effects on stellarator confinement in small devices

  19. Experimental investigations of driven Alfven wave resonances in a tokamak plasma using carbon dioxide laser interferometry

    International Nuclear Information System (INIS)

    The first direct observation of the internal structure of driven global Alfven eigenmodes in a tokamak plasma is presented. A carbon dioxide laser scattering/interferometer has been designed, built, and installed on the PRETEXT tokamak. By using this diagnostic system in the interferometer configuration, we have for the first time, thoroughly investigated the resonance conditions required for, and the spatial wave field structure of, driven plasma eigenmodes at frequencies below the ion cyclotron frequency in a confined, high temperature, tokamak plasma

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-11-01

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

  2. Stability analysis of tokamak plasmas; Analyse de stabilite de plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bourdelle, C

    2000-10-01

    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)

  3. A novel approach to linearization of the electromagnetic parameters of tokamaks with an iron core

    International Nuclear Information System (INIS)

    The equivalent model of an iron core tokamak is developed, in which the electromagnetic parameters of several pairs of coils in opposite series (PCOS) are not dependent on the saturation of the iron core during tokamak operation. With this the electromagnetic parameters of all the coils in an iron core tokamak can be linearized, As an example, the electromagnetic parameters of Hefei Super-conductive Tokamak with iron core (HT-7) are linearized, and it is in good agreement with the experimental results. The linearization approach can be applied in real time plasma control and electromagnetic analysis

  4. The use of iron shims to reduce the toroidal field ripple in Tokamaks

    International Nuclear Information System (INIS)

    A study was performed to investigate the use of laminated ferromagnetic material, iron shims, to reduce the toroidal field, TF, ripple in Tokamaks. Fixed geometry shims can reduce the TF ripple and will behave linearly for Low Field Tokamaks with central fields under 2 Tesla. For High Field Tokamaks with central fields under 6 Tesla, fixed geometry shims can reduce the TF ripple but will have some non-linear behavior. Variable geometry shims can reduce the field ripple significantly with complete linearity for High Field Tokamaks

  5. A novel approach to linearization of the electromagnetic parameters of tokamaks with an iron core

    Energy Technology Data Exchange (ETDEWEB)

    Fu, P. E-mail: fupeng@mail.ipp.ac.cn; Liu, Z.Z.; Zou, J.H

    2002-05-01

    The equivalent model of an iron core tokamak is developed, in which the electromagnetic parameters of several pairs of coils in opposite series (PCOS) are not dependent on the saturation of the iron core during tokamak operation. With this the electromagnetic parameters of all the coils in an iron core tokamak can be linearized, As an example, the electromagnetic parameters of Hefei Super-conductive Tokamak with iron core (HT-7) are linearized, and it is in good agreement with the experimental results. The linearization approach can be applied in real time plasma control and electromagnetic analysis.

  6. Maintenance concept development for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs

  7. Operation of cryostat vacuum vessel of HT-7 superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)]. E-mail: yangyu@ipp.ac.cn; Su, M. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2006-11-15

    The superconducting tokamak HT-7 has been in operation for over 10 years. The safe and reliable operation of its cryostat vacuum vessel, which contains the superconducting coils is essential for each experimental run since the superconducting toroidal field coils are contained inside the vessel. In this paper, the operation is reviewed with the emphasis on the analysis on anomalous pressure rises and the corresponding solutions. It is shown that under close monitoring and timely handling, the cryostat vacuum vessel could still satisfy the requirements of the experimental operation despite of the material aging. This provides guideline for vacuum operating of HT-7. The experiences should be valuable for other superconducting projects as well, including a whole superconducting tokamak under construction, EAST.

  8. On the economic prospects of nuclear fusion with tokamaks

    Science.gov (United States)

    Pfirsch, D.; Schmitter, K. H.

    1987-12-01

    A method of cost and construction energy estimation for tokamak fusion power stations conforming to the present stage of fusion development is described. The method is based on first-wall heat load constraints rather than Beta limitations, which, however, might eventually be the more critical of the two. It is used to discuss the economic efficiency of pure fusion, with particular reference to the European study entitled Environmental Impact and Economic Prospects of Nuclear Fusion (1986). It is shown that the claims made therein for the economic prospects of pure fusion with tokamaks, when discussed on the basis of the present-day technology, do not stand up to critical examination. A fusion-fission hybrid, however, could afford more positive prospects. Support for the stated method is derived when it is properly applied for cost estimation of advanced gas-cooled and Magnox reactors, the two examples presented by the European study to disprove it.

  9. Pulse length assessment of compact ignition tokamak designs

    International Nuclear Information System (INIS)

    A time-dependent zero-dimensional code has been developed to assess the pulse length and auxiliary heating requirements of Compact Ignition Tokamak (CIT) designs. By taking a global approach to the calculation, parametric studies can be easily performed. The accuracy of the procedure is tested by comparing with the Tokamak Simulation Code which uses theory-based thermal diffusivities. A series of runs is carried out at various levels of energy confinement for each of three possible CIT configurations. It is found that for cases of interest, ignition or an energy multiplication factor Q /approxreverse arrowgt/ 7 can be attained within the first half of the planned five-second flattop with 10--40 MW of auxiliary heating. These results are supported by analytic calculations. 18 refs., 7 figs., 2 tabs

  10. 3D MHD disruptions simulations of tokamaks plasmas

    Science.gov (United States)

    Paccagnella, Roberto; Strauss, Hank; Breslau, Joshua

    2008-11-01

    Tokamaks Vertical Displacement Events (VDEs) and disruptions simulations in toroidal geometry by means of a single fluid visco-resistive magneto-hydro-dynamic (MHD) model are presented in this paper. The plasma model, implemented in the M3D code [1], is completed with the presence of a 2D homogeneous wall with finite resistivity. This allows the study of the relatively slowly growing magneto-hydro-dynamical perturbation, the resistive wall mode (RWM), which is, in this work, the main drive of the disruptions. Amplitudes and asymmetries of the halo currents pattern at the wall are also calculated and comparisons with tokamak experimental databases and predictions for ITER are given. [1] W. Park, E.V. Belova, G.Y. Fu, X.Z. Tang, H.R. Strauss, L.E. Sugiyama, Phys. Plasmas 6 (1999) 1796.

  11. X-ray tomography on IR-T1 tokamak

    International Nuclear Information System (INIS)

    Measurements of soft X-ray emission from the IR-T1 Tokamak plasma have been analyzed with tomographic reconstruction methods. The detector system is on a circular arc view IR-T1 Tokamak plasmas through a thin beryllium window to provide poloidal imaging of soft X-ray emission (SXR) in the 600 ev--20 Kev range. Each detector has an active area 3 x 20 mm2. The signals are digitized every 4 micro or 8 micros by use of 8 bit transient recorders and the frequency response of detector--preamplifier system is from 10 Hz--200 kHz. This method has been used to study the internal disruptions and m = 1 oscillation. The reconstructed images in the phase before a Sawtooth crash have shown the hot plasma core move gradually toward one side, this is considered that as the increasing m = 1 kink mode

  12. Equilibrium calculations for plasma control in CIT [Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in design and analysis of the poloidal field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for TSC. In this report, VEQ and its role in the TSC analysis of the CIT PF system are described. Equilibrium and coil current calculations are discussed, an overview of the CIT PF system is presented, a set of reference equilibria for modeling a complete discharge in CIT is described, and the concept of a plasma shape control matrix is introduced. 9 refs., 8 figs., 7 tabs

  13. Analytical solutions for Tokamak equilibria with reversed toroidal current

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Caroline G. L.; Roberto, M.; Braga, F. L. [Departamento de Fisica, Instituto Tecnologico de Aeronautica, Sao Jose dos Campos, Sao Paulo 12228-900 (Brazil); Caldas, I. L. [Instituto de Fisica, Universidade de Sao Paulo, 05315-970 Sao Paulo, SP (Brazil)

    2011-08-15

    In tokamaks, an advanced plasma confinement regime has been investigated with a central hollow electric current with negative density which gives rise to non-nested magnetic surfaces. We present analytical solutions for the magnetohydrodynamic equilibria of this regime in terms of non-orthogonal toroidal polar coordinates. These solutions are obtained for large aspect ratio tokamaks and they are valid for any kind of reversed hollow current density profiles. The zero order solution of the poloidal magnetic flux function describes nested toroidal magnetic surfaces with a magnetic axis displaced due to the toroidal geometry. The first order correction introduces a poloidal field asymmetry and, consequently, magnetic islands arise around the zero order surface with null poloidal magnetic flux gradient. An analytic expression for the magnetic island width is deduced in terms of the equilibrium parameters. We give examples of the equilibrium plasma profiles and islands obtained for a class of current density profile.

  14. Commissioning of heating neutral beams for COMPASS-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Deichuli, P.; Davydenko, V.; Belov, V.; Gorbovsky, A.; Dranichnikov, A.; Ivanov, A.; Sorokin, A.; Mishagin, V.; Abdrashitov, A.; Kolmogorov, V.; Kondakov, A. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation)

    2012-02-15

    Two neutral beam injectors have been developed for plasma heating on COMPASS-D tokamak (Institute of Plasma Physics, Prague). The 4-electrodes multihole ion-optical system with beam focusing was chosen to provide the low divergence 300 kW power in both deuterium and hydrogen atoms. The accelerating voltage is 40 kV at extracted ion current up to 15 A. The power supply system provides the continuous and modulated mode of the beam injection at a maximal pulse length 300 ms. The optimal arrangement of the cryopanels and the beam duct elements provides sufficiently short-length beamline which reduces the beam losses. The evolution of the impurities and molecular fraction content is studied in the process of the high voltage conditioning of the newly made ion sources. Two injectors of the same type have been successfully tested and are ready for operation at tokamak in IPP, Prague.

  15. Tokamak resistive magnetohydrodynamic ballooning instability in the negative shear regime

    Institute of Scientific and Technical Information of China (English)

    Shi Bing-Ren; Lin Jian-Long; Li Ji-Quan

    2007-01-01

    Improved confinement of tokamak plasma with central negative shear is checked against the resistive ballooning mode. In the negative shear regime, the plasma is always unstable for purely growing resistive ballooning mode. For a simplest tokamak equilibrium model, the s-α model, characteristics of this kind of instability are fully clarified by numerically solving the high n resistive magnetohydrodynamic ballooning eigen-equation. Dependences of the growth rate on the resistivity, the absolute shear value, the pressure gradient are scanned in detail. It is found that the growth rate is a monotonically increasing function of a while it is not sensitive to the changes of the shear s, the initial phase θ0 and the resistivity parameter εR.

  16. On the Production of Relativistic Runaway Electrons in Damavand Tokamak

    Science.gov (United States)

    Moslehi-Fard, Mahmoud

    2013-02-01

    Experimental observations in Damavand tokamak show that hard X-ray is produced by either disruption with I p 20 kA. Hard X-ray also persists from the initiation of plasma discharge to the end. Occurrence of multiple spikes in hard X-ray during the discharge is evident. The propagation of hard X-ray is attributed to runaway electrons. We observe runaway electrons in two regimes with different characteristics. Regime (RADI) is similar to the observations of other Tokamak during disruption on that the plasma current is reduced abruptly and interpreted by Dreicer theory. In the regime of RADII, hard X-ray and subsequently runaway electrons are observed from starting of plasma discharge which provides the condition that the most of runaway electrons contain the toroidal plasma current. Runaway electron beam excites whistler waves and scattered electrons in velocity space and prevent growing the runaway electrons beam.

  17. On the economic prospects of nuclear fusion with tokamaks

    International Nuclear Information System (INIS)

    This paper describes a method of cost and construction energy estimation for tokamak fusion power stations conforming to the present, early stage of fusion development. The method is based on first-wall heat load constraints rather than β limitations, which, however, might eventually be the more critical of the two. It is used to discuss the economic efficiency of pure fusion, with particular reference to the European study entitled 'Environmental Impact and Economic Prospects of Nuclear Fusion'. It is shown that the claims made therein for the economic prospects of pure fusion with tokamaks, when discussed on the basis of the present-day technology, do not stand up to critical examination. A fusion-fission hybrid, however, could afford more positive prospects. Support for the stated method is even derived when it is properly applied for cost estimation of advanced gascooled and Magnox reactors, the two very examples presented by the European study to 'disprove' it. (orig.)

  18. Impact of a poloidal divertor in ignition tokamak design

    International Nuclear Information System (INIS)

    System design studies were performed to assess the effect of assuming a poloidal divertor instead of a limiter as a means of impurity control for ignition tokamak configurations. Results show that for the nominal Tokamak Fusion Core Experiment (TFCX) device with superconducting TF coils, a feasible poloidal divertor configuration can be obtained without increasing the major radius. In the TFCX nominal copper TF coil device, however, field limits at the PF coils are exceeded when the effects of asymmetry associated with a poloidal divertor are included. It was found that a 12% increase in the major radius of this device is necessary to simultaneously satisfy the plasma-shaping requirements of a poloidal divertor and the magnetics constraints at the superconducting PF coils

  19. Recent Progress of HT-7U Superconducting Tokamak

    Science.gov (United States)

    Weng, Pei-de

    2002-12-01

    HT-7U is a superconducting tokamak, which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak. The engineering design of the device has been optimized. The R&D program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.

  20. Imaging System and Plasma Imaging on HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    郑银甲; 冯震; 罗萃文; 刘莉; 李伟; 严龙文; 杨青巍; 刘永

    2004-01-01

    As a new diagnostic means, plasma-imaging system has been developed on the HL2A tokamak, with a basic understanding of plasma discharge scenario of the entire torus, checking the plasma position and the clearance between the plasma and the first wall during discharge. The plasma imaging system consists of (1) color video camera, (2) observation window and turn mirror,(3) viewing & collecting optics, (4) video cable, (5) Video capture card as well as PC. This paper mainly describes the experimental arrangement, plasma imaging system and detailed part in the system, along with the experimental results. Real-time monitoring of plasma discharge process,particularly distinguishing limitor and divertor configuration, the imaging system has become key diagnostic means and laid the foundation for further physical experiment on the HL-2A tokamak.

  1. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-15

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

  2. Performance projections for the Lithium Tokamak eXperiment (LTX)

    International Nuclear Information System (INIS)

    Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest enhancements in ohmic tokamak confinement ever observed. Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA code, utilizing a model with neoclassical ion transport and boundary conditions suitable to a nonrecycling wall, with fueling via edge gas puffing. This transport model has successfully reproduced the experimental values of the energy confinement (5-6 msec), loop voltage (2, during electron beam evaporation experiments on CDX-U. Similar results were obtained with solid surface overlaying a liquid lithium-filled target, although the solid surface was in this case composed of oxides of lithium. Here we reproduce the latter experiments, using a thin walled (0.125 mm) lithium-filled container, e-beam heated from the side. Results with various target geometries will be presented. (author)

  3. Transition to subcritical turbulence in a tokamak plasma

    CERN Document Server

    van Wyk, F; Schekochihin, A A; Roach, C M; Field, A R; Dorland, W

    2016-01-01

    Unstable perturbations driven by the pressure gradient and other sources of free energy in tokamak plasmas can grow exponentially and eventually saturate nonlinearly, leading to turbulence. Recent work has shown that in the presence of sheared flows, such systems can be subcritical. This means that all perturbations are linearly stable and a transition to a turbulent state only occurs if large enough initial perturbations undergo sufficient transient growth to allow nonlinear interaction. There is, however, currently very little known about a subcritical transition to turbulence in fusion-relevant plasmas. Here we use first-principles gyrokinetic simulations of a turbulent plasma in the outer core of the Mega-Ampere Spherical Tokamak (MAST) to demonstrate that the experimentally observed state is near the transition threshold, that the turbulence in this state is subcritical, and that transition to turbulence occurs via accumulation of very long-lived, intense, finite-amplitude coherent structures, which domi...

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

    International Nuclear Information System (INIS)

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

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

    CERN Document Server

    Casati, Alessandro

    2012-01-01

    The development of a quasi-linear gyrokinetic transport model for tokamak plasmas, ultimately designed to provide physically comprehensive predictions of the time evolution of the thermodynamic relevant quantities, is a task that requires tight links among theoretical, experimental and numerical studies. The framework of the model here proposed, which operates a reduction of complexity on the nonlinear self-organizing plasma dynamics, allows in fact multiple validations of the current understanding of the tokamak micro-turbulence. The main outcomes of this work stem from the fundamental steps involved by the formulation of such a reduced transport model, namely: (1) the verification of the quasi-linear plasma response against the nonlinearly computed solution, (2) the improvement of the turbulent saturation model through an accurate validation of the nonlinear codes against the turbulence measurements, (3) the integration of the quasi-linear model within an integrated transport solver.

  6. Magnetic surfaces of toroidal helical fields in tokamaks

    International Nuclear Information System (INIS)

    It is proposed to analyse theoretically the disruptive instability that occurs in confined plasmas in tokamaks, through investigation of the influence of resonant helical fields on their equilibrium; With this aim, a superposition of the magnetic field of the plasma in static MHD equilibrium with the field associated with the resonances is considered, taking into account the toroidal geometry of the tokamak. Due to the lack of symmetry, the lines of the total magnetic field resulting from this superposition must form magnetic surfaces only around some regions of the plasma. using the averaging method, functions of approximate magnetic surfaces are obtained (analytically) around the regions of resonances of the plasma (they contain the lines of this total magnetic field). It was verified that these approximate surfaces have structures of magnetic islands. (author)

  7. Statistical study of density fluctuations in the tore supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Devynck, P.; Fenzi, C.; Garbet, X.; Laviron, C. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Antar, G.; Gervais, F.; Hennequin, P.; Quemeneur, A.; Sabot, R.; Truc, A. [LPMI, CNRS UPR-287, Ecole Polytechnique, 91 - Palaiseau (France)

    1998-03-01

    It is believed that the radial anomalous transport in tokamaks is caused by plasma turbulence. Using infra-red laser scattering technique on the Tore Supra tokamak, statistical properties of the density fluctuations are studied as a function of the scales in ohmic as well as additional heating regimes using the lower hybrid or the ion cyclotron frequencies. The probability distributions are compared to a Gaussian in order to estimate the role of intermittency which is found to be negligible. The temporal behaviour of the three-dimensional spectrum is thoroughly discussed; its multifractal character is reflected in the singularity spectrum. The autocorrelation coefficient as well as their long-time incoherence and statistical independence. We also put forward the existence of fluctuations transfer between two distinct but close wavenumbers. A rather clearer image is thus obtained about the way energy is transferred through the turbulent scales. (author) 28 refs.

  8. Comparison of a radial fractional transport model with tokamak experiments

    Science.gov (United States)

    Kullberg, A.; Morales, G. J.; Maggs, J. E.

    2014-03-01

    A radial fractional transport model [Kullberg et al., Phys. Rev. E 87, 052115 (2013)], that correctly incorporates the geometric effects of the domain near the origin and removes the singular behavior at the outer boundary, is compared to results of off-axis heating experiments performed in the Rijnhuizen Tokamak Project (RTP), ASDEX Upgrade, JET, and DIII-D tokamak devices. This comparative study provides an initial assessment of the presence of fractional transport phenomena in magnetic confinement experiments. It is found that the nonlocal radial model is robust in describing the steady-state temperature profiles from RTP, but for the propagation of heat waves in ASDEX Upgrade, JET, and DIII-D the model is not clearly superior to predictions based on Fick's law. However, this comparative study does indicate that the order of the fractional derivative, α, is likely a function of radial position in the devices surveyed.

  9. Comparison of a radial fractional transport model with tokamak experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kullberg, A., E-mail: kulladam@ucla.edu; Morales, G. J.; Maggs, J. E. [Department of Physics and Astronomy, University of California, Los Angeles, Los Angeles, California 90095 (United States)

    2014-03-15

    A radial fractional transport model [Kullberg et al., Phys. Rev. E 87, 052115 (2013)], that correctly incorporates the geometric effects of the domain near the origin and removes the singular behavior at the outer boundary, is compared to results of off-axis heating experiments performed in the Rijnhuizen Tokamak Project (RTP), ASDEX Upgrade, JET, and DIII-D tokamak devices. This comparative study provides an initial assessment of the presence of fractional transport phenomena in magnetic confinement experiments. It is found that the nonlocal radial model is robust in describing the steady-state temperature profiles from RTP, but for the propagation of heat waves in ASDEX Upgrade, JET, and DIII-D the model is not clearly superior to predictions based on Fick's law. However, this comparative study does indicate that the order of the fractional derivative, α, is likely a function of radial position in the devices surveyed.

  10. Molecular emission in the edge plasma of T-10 tokamak

    International Nuclear Information System (INIS)

    The experiments on recording molecular emission in the edge plasma of the T-10 tokamak are described. To obtain reliable spectra with sufficient spectral, temporal, and spatial resolution, the optical circuit is optimized for various experimental conditions. Typical spectra measured in two sections of the tokamak are shown. It is shown that, upon varying the parameters of the discharge, the molecular spectrum not only changes significantly in intensity but also undergoes a qualitative change in the rotational and vibrational structure. For a detailed analysis, we use the Fulcher-α system (d3Πu–a3Σg+) of deuterium in the wavelength range from 590 to 640 nm. The rotational temperatures of ground state X1Σg+ and upper excited state d3Πu are estimated by the measured spectra

  11. Gyrokinetic Simulation of Global Turbulent Transport Properties in Tokamak Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Wang, W.X.; Lin, Z.; Tang, W.M.; Lee, W.W.; Ethier, S.; Lewandowski, J.L.V.; Rewoldt, G.; Hahm, T.S.; Manickam, J.

    2006-01-01

    A general geometry gyro-kinetic model for particle simulation of plasma turbulence in tokamak experiments is described. It incorporates the comprehensive influence of noncircular cross section, realistic plasma profiles, plasma rotation, neoclassical (equilibrium) electric fields, and Coulomb collisions. An interesting result of global turbulence development in a shaped tokamak plasma is presented with regard to nonlinear turbulence spreading into the linearly stable region. The mutual interaction between turbulence and zonal flows in collisionless plasmas is studied with a focus on identifying possible nonlinear saturation mechanisms for zonal flows. A bursting temporal behavior with a period longer than the geodesic acoustic oscillation period is observed even in a collisionless system. Our simulation results suggest that the zonal flows can drive turbulence. However, this process is too weak to be an effective zonal flow saturation mechanism.

  12. Total magnetic reconnection during a tokamak major disruption

    International Nuclear Information System (INIS)

    The safety factor within a tokamak plasma has been measured during a major disruption. During the disruption, the central safety factor jumps from below one to above one, while the total current is unchanged. This implies that total reconnection has occurred. This observation is in contract to the absence of total reconnection observed during a sawtooth oscillation in the same device. 11 refs., 6 figs

  13. Rotation, turbulence and transport in the Tokamak de Varennes

    International Nuclear Information System (INIS)

    This work was undertaken to achieve greater understanding of the process of transport in a plasma under magnetic confinement in a tokamak while measuring different characteristics (speed, poloidal and toroidal rotation, ionic temperature, local emissivity, etc.) of different ionic populations (carbon, oxygen, hydrogen, etc.). An attempt was made to establish at what point transport behaviour diverges from the neoclassical theory and becomes anomalous. (L.L.) (5 figs., 1 ref.)

  14. Measurement of the internal magnetic field structure of tokamak plasmas

    International Nuclear Information System (INIS)

    The first part of this article deals with the physical fundaments and technical aspects of this polarimetric measuring method, with its diagnostic capability, but also with its limitations. The second part summarizes the essential experimental results and their feedback on the theoretical description of Tokamak plasmas, which caused a revision of the accepted ideas of the magnetic field structure and its magnetohydrodynamic stability, in particular in the area of the hot plasma core. (orig.)

  15. Neutronics design for a spheric tokamak fusion-transmutation reactor

    International Nuclear Information System (INIS)

    Based on studies of spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. A set of plasma parameters suitable for the transmutation blanket is selected. Using the transport and burn-up calculation code BISON3.0 and its associated database, transmutation rate of MA nuclear waste, energy multiplication, and tritium breeder rate in the transmutation blanket are calculated

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

    International Nuclear Information System (INIS)

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

  17. HCN Laser Interferometer on the EAST Superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

    XU Qiang; GAO Xiang; JIE Yinxian; LIU Haiqing; SHI Nan; CHENG Yongfei; TONG Xingde

    2008-01-01

    A single-channel far-infrared (FIR) laser interferometer was developed to measure the line averaged electron density on the EAST tokamak. The structure of the single-channel FIR laser interferometer is described in detail. The evolution of density sawtooth oscillation was measured by means the FIR laser interferometer, and was identified by electron cyclotron emission (ECE) signals and soft X-ray intensity. The discharges with and without sawtooth were compared with each other in the Hugill diagram.

  18. Correlations of heat and momentum transport in the TFTR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Scott, S. D.; Arunasalam, V.; Barnes, C. W.; Bell, M. G.; Bitter, M.; Boivin, R.; Bretz, N. L.; Budny, R.; Bush, C. E.; Cavallo, A.; Chu, T. K.; Cohen, S. A.; Colestock, P.; Davis, S. L.; Dimock, D. L.; Dylla, H. F.; Efthimion, P. C.; Erhrardt, A. B.; Fonck, R. J.; Fredrickson, E.; Furth, H. P.; Goldston, R. J.; Greene, G.; Grek, B.; Grisham, L. R.; Hammett, G.; Hawryluk, R. J.; Hendel, H. W.; Hill, K. W.; Hinnov, E.; Hoffman, D. J.; Hosea, J.; Howell, R. B.; Hsuan, H.; Hulse, R. A.; Jaehnig, K. P.; Janos, A. C.; Jassby, D.; Jobes, F.; Johnson, D. W.; Johnson, L. C.; Kaita, R.; Kieras-Phillips, C.; Kilpatrick, S. J.; LaMarche, P. H.; LeBlanc, B.; Little, R.; Manos, D. M.; Mansfield, D. K.; Mazzucato, E.; McCarthy, M. P.; McCune, D. C.; McGuire, K.; McNeill, D. H.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Motley, R.; Mueller, D.; Murphy, J. A.; Nagayama, Y.; Nazakian, R.; Owens, D. K.; Park, H.; Ramsey, A. T.; Redi, M. H.; Roquemore, A. L.; Rutherford, P. H.; Schilling, G.; Schivell, J.; Schmidt, G. L.; Stevens, J.; Stratton, B. C.; Stodiek, W.; Synakowski, E. J.; Tang, W. M.; Taylor, G.; Timberlake, J. R.; Towner, H. H.; Ulrickson, M.; von Goeler, S.; Wieland, R.; Williams, M.; Wilson, J. R.; Wong, K.-L.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zweben, S. J.

    1990-01-01

    Measurements of the toroidal rotation speed v φ (r) driven by neutral beam injection in tokamak plasmas and, in particular, simultaneous profile measurements of v φ, Ti, Te, and ne, have provided new insights into the nature of anomalous transport in tokamaks. Low-recycling plasmas heated with unidirectional neutral beam injection exhibit a strong correlation among the local diffusivities, χφ≈χi}>χe. Recent measurements have confirmed similar behavior in broad-density L-mode plasmas. These results are consistent with the conjecture that electrostatic turbulence is the dominant transport mechanism in the tokamak fusion test reactor tokamak (TFTR) (Phys. Rev. Lett. 58, 1004 (1987)), and are inconsistent with predictions both from test-particle models of strong magnetic turbulence and from ripple transport. Toroidal rotation speed measurements in peaked-density TFTR supershots'' with partially unbalanced beam injection indicate that momentum transport decreases as the density profile becomes more peaked. In high-temperature, peaked-density plasmas the observed gradient scale length parameter η tot i=d ln T i/d ln n e correlates reasonably well with predictions of the threshold for exciting ion-temperature-gradient-driven turbulence (ITGDT), as would be expected for plasmas at marginal stability with respect to this strong transport mechanism. In L-mode plasmas where ITGDT is expected to be too weak to enforce marginal stability, ηtoti exceeds this threshold considerably.

  19. Design of geometric phase measurement in EAST Tokamak

    CERN Document Server

    Lan, T; Liu, J; Jie, Y X; Wang, Y L; Gao, X; Qin, H

    2016-01-01

    The optimum scheme for geometric phase measurement in EAST Tokamak is proposed in this paper. The theoretical values of geometric phase for the probe beams of EAST Polarimeter-Interferometer (POINT) system are calculated by path integration in parameter space. Meanwhile, the influences of some controllable parameters on geometric phase are evaluated. The feasibility and challenge of distinguishing geometric effect in the POINT signal are also assessed in detail.

  20. FEM [Free Electron Maser] for tokamak: Final report

    International Nuclear Information System (INIS)

    This paper studies the feasibility of a microwave source for heating a tokamak reactor. The free electron maser (FEM) shows great promise for being this source. The topics covered in this paper are microwave generation with FEM, efficiency enhancement, parameter scaling, space charge scaling, beam energy spread and efficiency scaling, electron beam line with energy recovery, achromatic bend, multi-stage depressed voltage electron beam collector, and development plans. 12 refs., 10 figs., 5 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Castracane, J.

    2001-01-04

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

  2. Rotational soft x-ray tomography of noncircular tokamak plasmas

    International Nuclear Information System (INIS)

    A rotational tomography technique for noncircular tokamak plasmas has been developed. Using a linear transformation from an elliptic coordinate system to the circular one, and compensating for the Shafranov shift, the elliptic plasma shape is transformed to the concentric circular shape. Fitting the data of a quarter rotation to the Fourier--Bessel expansions, the tomography is performed. This technique is applied to the snake oscillation, to the slow sawtooth crash, and to the post-cursor oscillations of noncircular plasmas on JET

  3. Soft x-ray tomography on HT-7 tokamak

    International Nuclear Information System (INIS)

    The hardware of the soft x-ray imaging system on HT-7 tokamak is described. Fourier-Bessel expansion method is used to reconstruct the soft x-ray emission distribution. The simulation results show that the inversion accuracy is associated with the angular expansion limit and the hot core displacement. Besides, tomographic results of sawtooth oscillations in a typical LHCD plasma are presented. (author)

  4. Interactive exploration of tokamak turbulence simulations in virtual reality

    International Nuclear Information System (INIS)

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

  5. First 50 pps Thomson scattering diagnostics in a tokamak

    International Nuclear Information System (INIS)

    Electron temperature and density measurements by Thomson scattering were performed for the first time for the whole duration of a tokamak discharge. A 50 pps Nd:YAG laser at 1.06 μm was used in ASDEX in combination with Si avalanche photodiode detectors. Density calibration was done by rotational anti-Stokes Raman scattering from hydrogen. The system is used for measurements at electron densities of as low as 2 x 1012 cm-3. (orig.)

  6. Structural materials for large superconducting magnets for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Long, C.J.

    1976-12-01

    The selection of structural materials for large superconducting magnets for tokamak-type fusion reactors is considered. The important criteria are working stress, radiation resistance, electromagnetic interaction, and general feasibility. The most advantageous materials appear to be face-centered-cubic alloys in the Fe-Ni-Cr system, but high-modulus composites may be necessary where severe pulsed magnetic fields are present. Special-purpose structural materials are considered briefly.

  7. A Steady State Tokamak Operation by Use of Magnetic Monopoles

    OpenAIRE

    Narihara, K.

    1991-01-01

    A steady state tokamak operation based on a magnetic monopole circuit is considered. Circulation of a chain of iron cubes which trap magnetic monopoles generates the needed loop voltage. The monopole circuit is enclosed by a series of solenoid coils in which the magnetic field is feedback controlled so that the force on the circuit balance against the mechanical friction. The driving power is supplied through the current sources of poloidal, ohmic and solenoid coils. The current drive efficie...

  8. Mode Analysis with Autocorrelation Method (Single Time Series) in Tokamak

    Science.gov (United States)

    Saadat, Shervin; Salem, Mohammad K.; Goranneviss, Mahmoud; Khorshid, Pejman

    2010-08-01

    In this paper plasma mode analyzed with statistical method that designated Autocorrelation function. Auto correlation function used from one time series, so for this purpose we need one Minov coil. After autocorrelation analysis on mirnov coil data, spectral density diagram is plotted. Spectral density diagram from symmetries and trends can analyzed plasma mode. RHF fields effects with this method ate investigated in IR-T1 tokamak and results corresponded with multichannel methods such as SVD and FFT.

  9. Studies on fundamental technologies for producing tokamak-plasma

    International Nuclear Information System (INIS)

    The report describes studies on fundamental technologies to produce tokamak-plasma of the JFT-2 and JFT-2M tokamaks. (1) In order to measure the particle number of residual gases, calibration methods of vacuum gauges have been developed. (2) Devices for a Taylor-type discharge cleaning (TDC), a glow discharge cleaning (GDC) and ECR discharge cleaning (ECR-DC) have been made and the cleaning effects have been investigated. In TDC the most effective plasma for cleaning is obtained in the plasma with 5 eV of electron temperature. GDC is effective in removing carbon impurities, but is less effective for removing oxygen impurities. ECR-DC has nearly the similar effect as TDC. The cleaning effect of these three types were studied by comparing the properties of resulting tokamak plasmas in the JFT-2M tokamak. (3) Experimental studies of pre-ionization showed as following results; A simple pre-ionization equipment as a hot-electron-gun and a J x B gun was effective in reducing breakdown voltage. An ordinary mode wave of the electron cyclotron frequency was very effective for pre-ionization. The RF power whose density is 3.6 x 10-2 W/cm3 produced plasma of an electron density of 5 x 1011 cm-3. In this case, it is possible to start up with negligible consumption of the magnetic flux caused by the plasma resistance. (4) Concerning to studies on plasma control, the following results were obtained; In order to obtain constant plasma current, a pulse forming network was constructed and sufficient constant plasma current was achieved. In applying an iso-flux method for measuring the plasma position, it is no problem practically to use only one loop-coil and one magnetic probe. (author)

  10. Design of Magnetic Measurement System on SUNIST Spherical Tokamak

    Institute of Scientific and Technical Information of China (English)

    ZENG Long; LIU Jun; WANG Ping; ZHANG Lu; HE Yexi; GAO Zhe; WANG Wenhao; XIE Lifeng; TAN Yi; ZHANG Liang; XIE Huiqiao; PENG Lili

    2008-01-01

    A magnetic measurement system consisting of magnetic probes and flux loops for spherical tokamak SUNIST,is uniquely designed due to the strongly shaped plasma cross section and the narrow space near the central solenoid.Plasma equilibrium reconstruction with the current filament method is performed to determine the number and positions of the magnetic probes and flux loops,as well as their design precision required.

  11. Experimental observation of the shear Alfven resonance in a tokamak

    International Nuclear Information System (INIS)

    Experiments in Tokapole II have demonstrated the shear Alfven resonance in a tokamak by direct probe measurement of the wave magnetic field within the plasma. The resonance is driven by external antennas and is identified as radially localized enhancements of the poloidal wave magnetic field. The radial location agrees with calculations which include toroidicity and noncircularity of the plasma cross-section. Other properties such as polarization, radial width, risetime, and wave enhancement also agree with MHD theory

  12. High Wavenumber Density Fluctuation Measurement in the HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    LI Yadong; LI Jiangang; ZHANG Xiaodong; ZHANG Tao; LIN Shiyao

    2009-01-01

    A three channel CO_2 laser forward scattering diagnostic system was set up to study the core plasma density fluctuation in the HT-7 tokamak.The range of the wavenumber measurement in the poloidal direction was kθ=12 cm~(-1) to 34 cm~(-1) and the wavenumber resolution was △k=2.5 cm~(-1).The wavenumber spectrum,the coexisting modes and the micro-instability eruption in experiments are presented.

  13. Development of Alfven wave antenna system for TCABR Tokamak

    International Nuclear Information System (INIS)

    The advanced antenna system for Alfven wave plasma heating and current drive in TCABR tokamak is presented. The antenna system is capable of exciting the travelling waves M=- 1, N=-4, -6 with single helicity and provides the possibility to improve Alfven wave plasma heating efficiency and to increase RF power input up 1 MW, without an uncontrolled density rise. The basic features of the antenna design and the results of preliminary tests are analyzed. (author)

  14. A classical picture of anomalous effects in a Tokamak

    Science.gov (United States)

    Hirano, K.

    1984-01-01

    Atomic collisions between plasma ions and a very small amount of neutral particles remaining in a hot plasma plays a very important role for plasma transports and may be an origin of anomalous effects observed in a Tokamak such as the diffusion coefficient independent of the field strength, a rapid plasma density increase during gas puffing and current penetration with anomalously high speed in the start-up phase. The Ohm's law derived by Cowling is used for the analysis.

  15. Multi-field plasma sandpile model in tokamaks and applications

    Science.gov (United States)

    Peng, X. D.; Xu, J. Q.

    2016-08-01

    A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.

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

    International Nuclear Information System (INIS)

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

  17. Development of tokamak reactor automated design code 'TRADE'

    International Nuclear Information System (INIS)

    This report describes Tokamak Reactor Automated Design Code ''TRADE'' which has been developed in order to assess the impact of the design choices on reactor systems and to find out the optimum design concept through wide parameter range. The optimum design concept is realized by finding out the optimum compromise between the plasma performance, the torus structure and the coil systems. Furthermore, TRADE code is for producing an input data set into TORSAC which had already been developed for a sensitivity analysis. (author)

  18. Control of the Vertical Instability in the SUNIST Spherical Tokamak

    Institute of Scientific and Technical Information of China (English)

    ZENG Long; GAO Zhe; HE Yexi

    2009-01-01

    Control of the vertical instability of the plasma in the SUNIST spherical tokamak is studied. A proportional-derivative (PD) position controller is employed in a control system which consists of a massless plasma, active coils and a passive conducting vessel. Stability conditions are determined and the feedback control with different combinations of both active and passive coils, available on SUNIST, is emphasized. The simulation results are also discussed.

  19. Imaging charge exchange recombination spectroscopy on the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Howard, J [Plasma Research Laboratory, The Australian National University, Canberra 0200 (Australia); Jaspers, R [Eindhoven University of Technology, Eindhoven (Netherlands); Lischtschenko, O; Delabie, E [FOM Institute for Plasma Physics ' Rijnhuizen' , Nieuwegein (Netherlands); Chung, J [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2010-12-15

    We describe the application of a simple spatial-heterodyne coherence-imaging filter for 2D Doppler imaging of charge exchange recombination (CXR) emission from a heating beam in the TEXTOR tokamak. Results obtained by the CXR imaging system are found to be consistent with measurements obtained using a standard multi-channel spectrometer-based system. We describe the system, indicate possible enhancements and future applications for imaging CXRS.

  20. Imaging charge exchange recombination spectroscopy on the TEXTOR tokamak

    Science.gov (United States)

    Howard, J.; Jaspers, R.; Lischtschenko, O.; Delabie, E.; Chung, J.

    2010-12-01

    We describe the application of a simple spatial-heterodyne coherence-imaging filter for 2D Doppler imaging of charge exchange recombination (CXR) emission from a heating beam in the TEXTOR tokamak. Results obtained by the CXR imaging system are found to be consistent with measurements obtained using a standard multi-channel spectrometer-based system. We describe the system, indicate possible enhancements and future applications for imaging CXRS.

  1. Dust-Particle Transport in Tokamak Edge Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K; Rognlien, T D

    2005-09-12

    Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.

  2. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb3Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered

  3. Modeling of Anomalous Transport in Tokamaks with FACETS code

    Science.gov (United States)

    Pankin, A. Y.; Batemann, G.; Kritz, A.; Rafiq, T.; Vadlamani, S.; Hakim, A.; Kruger, S.; Miah, M.; Rognlien, T.

    2009-05-01

    The FACETS code, a whole-device integrated modeling code that self-consistently computes plasma profiles for the plasma core and edge in tokamaks, has been recently developed as a part of the SciDAC project for core-edge simulations. A choice of transport models is available in FACETS through the FMCFM interface [1]. Transport models included in FMCFM have specific ranges of applicability, which can limit their use to parts of the plasma. In particular, the GLF23 transport model does not include the resistive ballooning effects that can be important in the tokamak pedestal region and GLF23 typically under-predicts the anomalous fluxes near the magnetic axis [2]. The TGLF and GYRO transport models have similar limitations [3]. A combination of transport models that covers the entire discharge domain is studied using FACETS in a realistic tokamak geometry. Effective diffusivities computed with the FMCFM transport models are extended to the region near the separatrix to be used in the UEDGE code within FACETS. 1. S. Vadlamani et al. (2009) %First time-dependent transport simulations using GYRO and NCLASS within FACETS (this meeting).2. T. Rafiq et al. (2009) %Simulation of electron thermal transport in H-mode discharges Submitted to Phys. Plasmas.3. C. Holland et al. (2008) %Validation of gyrokinetic transport simulations using %DIII-D core turbulence measurements Proc. of IAEA FEC (Switzerland, 2008)

  4. The design of the Tokamak Physics Experiment (TPX)

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, J.A.; Goldston, R.J.; Sinnis, J.C.; Bernabei, S.; Bialek, J.M.; Bronner, G.; Chen, S.J.; Chrzanowski, J.; Citrolo, J.; Dahlgren, F.

    1993-09-01

    The Tokamak Physics Experiment (TPX) is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a low-activation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation of US industry. Operation will begin with first plasma in the year 2000.

  5. TOKAMAK EQUILIBRIA WITH CENTRAL CURRENT HOLES AND NEGATIVE CURRENT DRIVE

    Energy Technology Data Exchange (ETDEWEB)

    CHU, M.S.; PARKS, P.B.

    2002-06-01

    OAK B202 TOKAMAK EQUILIBRIA WITH CENTRAL CURRENT HOLES AND NEGATIVE CURRENT DRIVE. Several tokamak experiments have reported the development of a central region with vanishing currents (the current hole). Straightforward application of results from the work of Greene, Johnson and Weimer [Phys. Fluids, 3, 67 (1971)] on tokamak equilibrium to these plasmas leads to apparent singularities in several physical quantities including the Shafranov shift and casts doubts on the existence of this type of equilibria. In this paper, the above quoted equilibrium theory is re-examined and extended to include equilibria with a current hole. It is shown that singularities can be circumvented and that equilibria with a central current hole do satisfy the magnetohydrodynamic equilibrium condition with regular behavior for all the physical quantities and do not lead to infinitely large Shafranov shifts. Isolated equilibria with negative current in the central region could exist. But equilibria with negative currents in general do not have neighboring equilibria and thus cannot have experimental realization, i.e. no negative currents can be driven in the central region.

  6. Study of a disruption mitigation method for tokamak plasmas

    International Nuclear Information System (INIS)

    Disruptions are a sudden loss of confinement of a tokamak plasma which take place in around 20 ms. They may lead to severe damaging of the tokamak structure, through heat deposition on Plasma Facing Components, electromagnetic stresses and relativistic runaway electrons. On future reactors, disruption mitigation will be critical. Massive gas injection is one of the methods proposed to mitigate disruptions. It was studied both experimentally and numerically in the thesis. Experiments on the Tore Supra and JET tokamaks showed that light gases (helium) were able to suppress runaway electrons. They induce a large density build-up which is large enough to suppress runaway production. On the contrary, heavier gases should be able to radiate more of the plasma thermal energy, but generate runaway electrons. All gases reduce electromagnetic forces. Gas mixtures have also been tested successfully to combine the advantages of the two types of gas. The gas jet penetration is linked to MHD instabilities enhancing the radial transport of the ionized gas, but preventing the neutrals from penetrating further inside a critical MHD surface. Massive gas injection simulations have been carried out using the 3D MHD code Jorek, by adding a neutral fluid model to the code. Results show that MHD instabilities are triggered more rapidly with high amounts of gas, and that successive rational surfaces are ergodized by the penetration of the density front in the plasma, in agreement with experimental observations. (author)

  7. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)

  8. Equilibrium reconstruction in the TCA/Br tokamak

    International Nuclear Information System (INIS)

    The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author)

  9. Self-consistent diverted tokamak equilibria with nonzero edge current

    Institute of Scientific and Technical Information of China (English)

    Shi Bing-Ren

    2012-01-01

    The semi-analytical method,previously used to construct model double-null and single-null diverted tokamak equilibria (Bingren Shi,Plasma Phys.Control Fusion 50 (2008) 085006,51 (2009) 105008,Nucl.Fusion 51 (2011) 023004),is extended to describe diverted tokamak equilibria with nonzero edge current,including the Pfirsch-Schlüter(PS) current.The PS current density is expressed in a way suitable to describe a diverted tokamak configuration in the near separatrix region.The model equilibrium is expressed by only two terms of the exact separable solutions of the GradShafranov equation,one of which is governed by a homogeneous ordinary differential equation,and the other by an inhomogeneous one.The particular merits of such a model configuration are that the internal region inside the separatrix and a suitable scrape-off layer can be simultaneously described by this exact solution.To investigate the physics in the region near the X-point,the magnetic surfaces can be satisfactorily described by approximate hyperbolic curves.

  10. Operation of a tokamak reactor in the radiative improved mode

    Science.gov (United States)

    Morozov, D. Kh.; Mavrin, A. A.

    2016-03-01

    The operation of a nuclear fusion reactor has been simulated within a model based on experimental results obtained at the TEXTOR-94 tokamak and other facilities in which quasistationary regimes were achieved with long confinement times, high densities, and absence of the edge-localized mode. The radiative improved mode of confinement studied in detail at the TEXTOR-94 tokamak is the most interesting such regime. One of the most important problems of modern tokamaks is the problem of a very high thermal load on a divertor (or a limiter). This problem is quite easily solved in the radiative improved mode. Since a significant fraction of the thermal energy is reemitted by an impurity, the thermal loading is significantly reduced. As the energy confinement time τ E at high densities in the indicated mode is significantly larger than the time predicted by the scaling of ITERH-98P(y, 2), ignition can be achieved in a facility much smaller than the ITER facility at plasma temperatures below 20 keV. The revealed decrease in the degradation of the confinement time τ E with an increase in the introduced power has been analyzed.

  11. Non-Axisymmetric Shaping of Tokamaks Preserving Quasi-Axisymmetry

    Energy Technology Data Exchange (ETDEWEB)

    Long-Poe Ku and Allen H. Boozer

    2009-06-05

    If quasi-axisymmetry is preserved, non-axisymmetric shaping can be used to design tokamaks that do not require current drive, are resilient to disruptions, and have robust plasma stability without feedback. Suggestions for addressing the critical issues of tokamaks can only be validated when presented with sufficient specificity that validating experiments can be designed. The purpose of this paper is provide that specificity for non-axisymmetric shaping. To our knowledge, no other suggestions for the solution of a number of tokamak issues, such as disruptions, have reached this level of specificity. Sequences of three-field-period quasi-axisymmetric plasmas are studied. These sequences address the questions: (1) What can be achieved at various levels of non-axisymmetric shaping? (2) What simplifications to the coils can be achieved by going to a larger aspect ratio? (3) What range of shaping can be achieved in a single experimental facility? The sequences of plasmas found in this study provide a set of interesting and potentially important configurations.

  12. Performance Projections For The Lithium Tokamak Experiment (LTX)

    Energy Technology Data Exchange (ETDEWEB)

    Majeski, R.; Berzak, L.; Gray, T.; Kaita, R.; Kozub, T.; Levinton, F.; Lundberg, D. P.; Manickam, J.; Pereverzev, G. V.; Snieckus, K.; Soukhanovskii, V.; Spaleta, J.; Stotler, D.; Strickler, T.; Timberlake, J.; Yoo, J.; Zakharov, L.

    2009-06-17

    Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized.

  13. Analysis of neutral hydrogenic emission spectra in a tokamak

    Science.gov (United States)

    Ko, J.; Chung, J.; Jaspers, R. J. E.

    2015-10-01

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

  14. Statistical analysis of first period of operation of FTU Tokamak; Analisi statistica del primo periodo di operazioni del Tokamak FTU

    Energy Technology Data Exchange (ETDEWEB)

    Crisanti, F.; Apruzzese, G.; Frigione, D.; Kroegler, H.; Lovisetto, L.; Mazzitelli, G.; Podda, S. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Energia

    1996-09-01

    On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted.

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

    International Nuclear Information System (INIS)

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

  16. Calculation about a modification to the toroidal magnetic field of the Tokamak Novillo. Part I; Calculo sobre una modificacion al campo magnetico toroidal del Tokamak Novillo. Parte I

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-15

    The charged particles that constitute the plasma in the tokamaks are located in magnetic fields that determine its behavior. The poloidal magnetic field of the plasma current and the toroidal magnetic field of the tokamak possess relatively big gradients, which produce drifts on these particles. These drifts are largely the cause of the continuous lost of particles and of energy of the confinement region. In this work the results of numerical calculations of a modification to the 'traditional' toroidal magnetic field that one waits it diminishes the drifts by gradient and improve the confinement properties of the tokamaks. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

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

    Science.gov (United States)

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

    2015-12-01

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

  19. Including collisions in gyrokinetic tokamak and stellarator simulations

    Energy Technology Data Exchange (ETDEWEB)

    Kauffmann, Karla

    2012-04-10

    Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal

  20. Comparison between 2D turbulence model ESEL and experimental data from AUG and COMPASS tokamaks

    DEFF Research Database (Denmark)

    Ondac, Peter; Horacek, Jan; Seidl, Jakub;

    2015-01-01

    In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtained...

  1. Plasma Shape and Current Control Simulation of HT-7U Tokamak

    Institute of Scientific and Technical Information of China (English)

    吴斌; 张澄

    2003-01-01

    This paper describes the discharge simulation of HT-7U tokamak plasma equilibriumand plasma current by solving MHD equations and surface average transport equations using anequilibrium evolution code. The simulated result shows the evolution of plasma parameter versustime .The simulated result can play an important role in the design of the plasma equilibrium andcontrol system of a tokamak.

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

  4. Protection of tokamak plasma facing components by a capillary porous system with lithium

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I. [JSC «Red Star» Moscow (Russian Federation); National Research Nuclear University MEPhI, Moscow (Russian Federation); Vertkov, A. [JSC «Red Star» Moscow (Russian Federation); Mirnov, S.; Lazarev, V. [SSC TRINITI, Troitsk, Moscow (Russian Federation)

    2015-08-15

    Highlights: • CPS filled with liquid lithium applied as structural material of PFC in tokamak. • Behavior of lithium CPS at normal and disruption condition has been studied. • Concept of closed loop of lithium circulation in tokamak chamber has been confirmed experimentally. - Abstract: Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

  5. Stability of Tokamak Equilibrium with Internal Transport Barrier against High-n MHD Ballooning Mode

    Institute of Scientific and Technical Information of China (English)

    SHI Bing-Ren; QU Wen-Xiao; DONG Jia-Qi

    2006-01-01

    @@ A new eigen-mode equation for the tokamak high-n (the toroidal mode number) ideal magnetohydrodynamic (MHD) ballooning mode in tokamak plasmas is derived to include the toroidal effects that are significant for stability of configurations with internal transport barriers (ITBs).

  6. Tokamak plasma self-organization and the possibility to have the peaked density profile in ITER

    NARCIS (Netherlands)

    Razumova, K. A.; Andreev, V. F.; Kislov, A. Y.; Kirneva, N. A.; Lysenko, S. E.; Pavlov, Y. D.; Shafranov, T. V.; Donne, A. J. H.; Hogeweij, G. M. D.; Spakman, G. W.; R. Jaspers,; Kantor, M.; Walsh, M.

    2009-01-01

    The self-organization of a tokamak plasma is a fundamental turbulent plasma phenomenon, which leads to the formation of a self-consistent pressure profile. This phenomenon has been investigated in several tokamaks with different methods of heating. It is shown that the normalized pressure profile ha

  7. The 2008 Public Release of the International Multi-tokamak Confinement Profile Database

    NARCIS (Netherlands)

    Roach, C. M.; Walters, M.; Budny, R. V.; Imbeaux, F.; Fredian, T. W.; Greenwald, M.; Stillerman, J. A.; Alexander, D. A.; Carlsson, J.; Cary, J. R.; Ryter, F.; Stober, J.; Gohil, P.; Greenfield, C.; Murakami, M.; Bracco, G.; Esposito, B.; Romanelli, M.; Parail, V.; Stubberfield, P.; Voitsekhovitch, I.; Brickley, C.; Field, A. R.; Sakamoto, Y.; Fujita, T.; Fukuda, T.; Hayashi, N.; Hogeweij, G. M. D.; Chudnovskiy, A.; Kinerva, N. A.; Kessel, C. E.; Aniel, T.; Hoang, G. T.; Ongena, J.; Doyle, E. J.; Houlberg, W. A.; Polevoi, A. R.

    2008-01-01

    This paper documents the public release PR08 of the International Tokamak Physics Activity (ITPA) profile database, which should be of particular interest to the magnetic confinement fusion community. Data from a wide variety of interesting discharges from many of the world's leading tokamak ex

  8. Controlled thermonuclear fusion in TOKAMAK type reactors, the European example: Joint European Torus (JET)

    International Nuclear Information System (INIS)

    The development of controlled thermonuclear reaction in TOKAMAK type reactors, and the main projects in the world are presented. The main characteristics of the JET (Joint European Torus) program, the perspectives for energy production, and the international cooperation for viable use of the TOKAMAK are analysed. (M.C.K.)

  9. Economic trends of tokamak power plants independent of physics scaling models

    International Nuclear Information System (INIS)

    This study examines the effects of plasma radius, field on axis, plasma impurity level, and aspect ratio on power level and unit capital cost, $/kW/sub e/, of tokamak power plants sized independent of plasma physics scaling models. It is noted that tokamaks sized in this manner are thermally unstable based on trapped particle scaling relationships. It is observed that there is an economic advantage for larger power level tokamaks achieved by physics independent sizing; however, the incentive for increased power levels is less than that for fission reactors. It is further observed that the economic advantage of these larger power level tokamaks is decreased when plasma thermal stability measures are incorporated, such as by increasing the plasma impurity concentration. This trend of economy with size obtained by physics independent sizing is opposite to that observed when the tokamak designs are constrained to obey the trapped particle and empirical scaling relationships

  10. Transport and turbulence in a magnetized plasma (application to tokamak plasmas); Transport et turbulence dans un plasma magnetise (application aux plasmas de tokamaks)

    Energy Technology Data Exchange (ETDEWEB)

    Sarazin, Y

    2004-03-01

    This document gathers the lectures made in the framework of a Ph.D level physics class dedicated to plasma physics. This course is made up of 3 parts : 1) collisions and transport, 2) transport and turbulence, and 3) study of a few exchange instabilities. More precisely the first part deals with the following issues: thermonuclear fusion, Coulomb collisions, particles trajectories in a tokamak, neo-classical transport in tokamaks, the bootstrap current, and ware pinch. The second part involves: particle transport in tokamaks, quasi-linear transport, resonance islands, resonance in tokamaks, from quasi to non-linear transport, and non-linear saturation of turbulence. The third part deals with: shift velocities in fluid theory, a model for inter-change instabilities, Rayleigh-Benard instability, Hasegawa-Wakatani model, and Hasegawa-Mima model. This document ends with a series of appendices dealing with: particle-wave interaction, determination of the curvature parameter G, Rossby waves.

  11. Extremely shaped plasmas to improve the Tokamak concept

    International Nuclear Information System (INIS)

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

  12. Engineering design for the HL-2M tokamak components

    International Nuclear Information System (INIS)

    Highlights: • The HL-2M is upgraded from HL-2A tokamak which is being constructed in SWIP. • The plasma major radius is 1.78 m, the minor radius is 0.65 m, and the plasma current will reach up to 3 MA with the elongation 1.6–1.8 and tri-angularity more than 0.5. • The new machine has been designed with demountable toroidal field coils and the vacuum vessel is an all-welded torus. • The coils, vacuum vessel and support structure are under fabricated in factories, and the construction of device will be finished in the end of next year. -- Abstract: The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current IP of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories

  13. Geodesic acoustic mode in tokamaks: local consideration and eigenvalue analysis

    International Nuclear Information System (INIS)

    A set of magnetohydrodynamic equations describing the geodesic acoustic mode (GAM) in tokamak plasmas is derived. The obtained equations take into account the presence of the energetic ions and allow to study energetic-ion-driven GAM instability perturbatively or non-perturbatively (EGAM mode). They are applicable to plasmas with β-bar q2≲1, where β-bar =βs/(1+βs), βs=cs2/vA2, cs is the sound velocity, vA is the Alfvén velocity, q is the tokamak safety factor. Using these equations, GAM/EGAM instability is studied in a local approach and by means of the eigenvalue analysis. It is shown that β-coupling (the coupling of Fourier harmonics of the perturbation due to finite β—ratio of the plasma pressure to the magnetic field pressure—and the curvature of the field lines) can be responsible for the radial structure of the GAM-mode. A conclusion is drawn that conditions for the GAM/EGAM instability to arise are mildest in the case of counter-injection of energetic ions with pitch angles χ2 < 0.6 and large ratio of Larmor radius of the energetic ions to a characteristic length of inhomogeneity of these ions. A numerical code solving the derived equations is developed. Specific calculations are carried out for tokamaks with a non-monotonic safety factor. On the other hand, it is found that due to the presence of the energetic ions the GAM/EGAM continuum can have an extremum even when the safety factor q(r) is monotonic, which indicates that global modes can exist also in this case. (paper)

  14. Triangularity effects on the collisional diffusion for elliptic tokamak plasma

    International Nuclear Information System (INIS)

    In this conference the effect of ellipticity and triangularity will be analyzed for axisymmetric tokamak in the collisional regime. Analytic forms for the magnetic field cross sections are taken from those derived recently by other authors [1,2]. Analytical results can be obtained in elliptic plasmas with triangularity by using an special system of tokamak coordinates recently published [3-5]. Our results show that triangularities smaller than 0.6, increases confinement for ellipticities in the range 1.2 to 2. This behavior happens for negative and positive triangularities; however this effect is stronger for positive than for negative triangularities. The maximum diffusion velocity is not obtained for zero triangularity, but for small negative triangularities. Ellipticity is also very important in confinement, but the effect of triangularity seems to be more important. High electric inductive field increases confinement, though this field is difficult to modify once the tokamak has been built. The analytic form of the current produced by this field is like that of a weak Ware pinch with an additional factor, which weakens the effect by an order of magnitude. The dependence of the triangularity effect with the Shafranov shift is also analyzed. References 1. - L. L. Lao, S. P. Hirshman, and R. M. Wieland, Phys. Fluids 24, 1431 (1981) 2. - G. O. Ludwig, Plasma Physics Controlled Fusion 37, 633 (1995) 3. - P. Martin, Phys. Plasmas 7, 2915 (2000) 4. - P. Martin, M. G. Haines and E. Castro, Phys. Plasmas 12, 082506 (2005) 5. - P. Martin, E. Castro and M. G. Haines, Phys. Plasmas 12, 102505 (2005)

  15. Ambipolarity in a tokamak with magnetic field ripple

    Science.gov (United States)

    Hazeltine, R. D.

    2016-08-01

    In view of the recognized importance of electrostatic fields regarding turbulent transport, the radial electric field in a tokamak with magnetic field ripple is reconsidered. Terms in the ambipolarity condition involving the radial derivative of the field are derived from an extended drift-kinetic equation, including effects of second order in the gyroradius. Such terms are of interest in part because of their known importance in rotational relaxation equations for the axisymmetric case. The electric field is found to satisfy a nonlinear differential equation that is universal in a certain sense, and that implies spatial relaxation of the potential to its conventionally predicted value.

  16. Overview of tritium systems for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is being designed at several laboratories to produce and study fully ignited plasma discharges. The tritium systems which will be needed for CIT include fueling systems and radiation monitoring and safety systems. Design of the tritium systems is the responsibility of the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. Major new tritium systems for CIT include a pellet injector, an air detritiation system and a glovebox atmosphere detritiation system. The pellet injector is being developed at Oak Ridge National Laboratory. 7 refs., 2 figs

  17. Double adiabatic theory of collisionless geodesic acoustic modes in tokamaks

    CERN Document Server

    Hassam, A B

    2011-01-01

    Collisionless geodesic acoustic modes in tokamaks being supersonic for large "safety factor" q, the CGL (G. Chew, M. Goldberger, F. Low, 1956)1 double-adiabatic fluid closure is applied to formulate a theory for these modes. The basic linear normal mode is obtained. External means to drive these modes at resonance, as has been proposed earlier, are explored. The external drivers considered include external magnetic forces to effect flux surface displacements, as well as non-axisymmetric ion heating. Finally, the damping of these modes from collisional magnetic pumping is investigated using a model set of CGL collision-corrected equations.

  18. Halo current diagnostic system of experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chen, D. L.; Shen, B.; Sun, Y.; Qian, J. P., E-mail: jpqian@ipp.ac.cn; Wang, Y.; Xiao, B. J. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031 (China); Granetz, R. S. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States)

    2015-10-15

    The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

  19. Negative edge plasma currents in the SINP tokamak

    Indian Academy of Sciences (India)

    Ramesh Narayanan; A N Sekar Iyengar

    2011-12-01

    A tokamak plasma discharge having an increase in duration accompanied with enhanced runaway electron flux has been experimentally studied in this paper. The discharges have been obtained by controlling the applied vertical magnetic field ($B^{\\text{appl}}_v$) to below a critical value. Such discharges have been observed to have ‘negative edge plasma currents’, detected using an internal Rogowskii coil (IRC). We have tried to correlate the runaway behaviour with the negative edge plasma currents and have explained that these observations are a result of beam plasma instabilities.

  20. Characterization of the Tokamak de Varennes ohmic plasma in equilibrium

    International Nuclear Information System (INIS)

    Experimental results obtained on the Tokamak de Varennes during a series of reproducible ohmic discharges are presented. Profiles of basic plasma parameters are constructed and compared with theoretical predictions. In particular, the measured plasma resistivity agrees with the neoclassical scaling rather than with Spitzer resistivity. The study of electron-density fluctuations indicates a linear dispersion relation with a propagation velocity of 3.0 x 104 cm s-1. Particle transport investigations are initiated, giving experimental diffusion and convection coefficients across the plasma radius for electrons and impurity ions. (Author) 32 refs., 10 figs

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

    Directory of Open Access Journals (Sweden)

    Vlainić Miloš

    2015-06-01

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

  2. A midsize tokamak as a fast track to burning plasmas

    Directory of Open Access Journals (Sweden)

    E. Mazzucato

    2011-03-01

    Full Text Available This paper describes the conceptual design of a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥ 10 with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER. This can be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a different magnetic divertor from those currently employed in present experiments is discussed.

  3. Tokamak plasma shape identification based on the boundary integral equations

    International Nuclear Information System (INIS)

    A necessary condition for tokamak plasma shape identification is discussed and a new identification method is proposed in this article. This method is based on the boundary integral equations governing a vacuum region around a plasma with only the measurement of either magnetic fluxes or magnetic flux intensities. It can identify various plasmas with low to high ellipticities with the precision determined by the number of the magnetic sensors. This method is applicable to real-time control and visualization using a 'table-look-up' procedure. (author)

  4. Review of the Equilibrium Fitting for Non-Circular Tokamak

    Institute of Scientific and Technical Information of China (English)

    罗家融

    2002-01-01

    As the equilibrium fitting code (EFIT) is developing to perform the magnetic and the kinetic-magnetic analysis for tokamak device operation, it can be not only run in either the fitting mode or the equilibrium mode but also control operation of modern experimental fusion device. In this paper the history of EF1T code and its capabilities are described in section 2. A brief description of the off-line EFIT code and the development of the real-time EFIT (RTEFIT)code is shown in section 3 and 4 respectively. In the last section the summary of this paper is given.

  5. Hamiltonian four-field model for nonlinear tokamak dynamics

    International Nuclear Information System (INIS)

    The Hamiltonian four-field model is a simplified description of nonlinear tokamak dynamics that allows for finite ion Larmor radius physics as well as other effects related to compressibility and electron adiabaticity. Much simpler than a rigorous or even reduced description of the same physics, it still preserves essential features of the underlying exact dynamics. In particular, because it is a Hamiltonian dynamical system it conserves the appropriate Casimir invariants, as well as avoiding implicit, unphysical dissipation. Here the model is derived and interpreted, its Hamiltonian nature is demonstrated, and its constants of motion are extracted

  6. Global energy balance and density limit on CASTOR tokamak

    International Nuclear Information System (INIS)

    Total radiative power losses were measured with a pyroelectric detector on the CASTOR tokamak in a broad range of plasma parameters. It was shown that while the most important channel of energy losses for the low density operation (average ne19 m-3) is thermal conductivity, the high density regimes are radiative dominant. Using a simple analytic energy balance model, the connection is discussed between such a high level of radiation and the shrinking of the current channel resulting in an enhanced MHD activity. (author)

  7. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  8. Soft x-ray tomography on the RTP tokamak

    International Nuclear Information System (INIS)

    A soft x-ray tomography system is being constructed for the RTP tokamak. The main aim of this system is to measure fluctuations in the plasma emissivity with relatively high poloidal m number. Five pinhole cameras will be distributed inside the vessel in such a way that it becomes possible to reconstruct modes up to m=4, without making any assumption about the plasma rotation. The system has been designed to obtain a minimum bandwidth of 200 kHz and a temporal resolution of 1 μs. We describe the hardware that is under construction

  9. Tokamak with in situ magnetohydrodynamic generation of toroidal magnetic field

    Science.gov (United States)

    Schaffer, Michael J.

    1986-01-01

    A tokamak apparatus includes an electrically conductive metal pressure vessel for defining a chamber and confining liquid therein. A liner disposed within said chamber defines a toroidal space within the liner and confines gas therein. The metal vessel provides an electrically conductive path linking the toroidal space. Liquid metal is forced outwardly through the chamber outside of the toroidal space to generate electric current in the conductive path and thereby generate a toroidal magnetic field within the toroidal space. Toroidal plasma is developed within the toroidal space about the major axis thereof.

  10. Experimental measurement of electron heat diffusivity in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Callen, J.D.; Jahns, G.L.

    1976-06-01

    The electron temperature perturbation produced by internal disruptions in the center of the Oak Ridge Tokamak (ORMAK) is followed with a multi-chord soft x-ray detector array. The space-time evolution is found to be diffusive in character, with a conduction coefficient larger by a factor of 2.5 - 15 than that implied by the energy containment time, apparently because it is a measurement for the small group of electrons whose energies exceed the cut-off energy of the detectors.

  11. Application of toroidal multipoles to facilitate tokamak reactor studies

    International Nuclear Information System (INIS)

    A method of calculating the vacuum magnetic flux for plasma equilibrium over a range of aspect ratio A and major radius R is presented. Toroidal multipoles are used to fit the vacuum flux for a set of reference equilibria and the fitted moments are then used to parametrize the flux over the design space in A and R. An example is given in which the equilibrium flux for a D-shaped tokamak plasma is predicted and compared with the actual equilibrium code vacuum flux

  12. Divertor bypass in the Alcator C-Mod tokamak

    Science.gov (United States)

    Pitcher, C. S.; LaBombard, B.; Danforth, R.; Pina, W.; Silveira, M.; Parkin, B.

    2001-01-01

    The Alcator C-Mod divertor bypass has for the first time allowed in situ variations to the mechanical baffle design in a tokamak. The design utilizes small coils which interact with the ambient magnetic field inside the vessel to provide the torque required to control small flaps of a Venetian blind geometry. Plasma physics experiments with the bypass have revealed the importance of the divertor baffling to maintain high divertor gas pressures. These experiments have also indicated that the divertor baffling has only a limited effect on the main chamber pressure in C-Mod.

  13. Experimental study of the principles governing tokamak transport

    Science.gov (United States)

    Wagner, F.; Gruber, O.; Lackner, K.; Murmann, H. D.; Speth, E.; Becker, G.; Bosch, H. S.; Brocken, H.; Cattanei, G.; Dorst, D.; Eberhagen, A.; Elsner, A.; Erckmann, V.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Grieger, G.; Grigull, P.; Haas, G.; Hacker, H.; Hartfuss, H. J.; Jäckel, H.; Jaenicke, R.; Janeschitz, G.; Junker, J.; Karger, F.; Kasparek, W.; Keilhacker, M.; Kick, M.; Klüber, O.; Kornherr, M.; Kroiss, H.; Kuehner, M.; Lenoci, M.; Lisitano, G.; Maassberg, M.; Mahn, C.; Marlier, S.; Mayer, H. M.; McCormick, K.; Meisel, D.; Mertens, V.; Müller, E. R.; Müller, .; Müller, G.; Niedermeyer, H.; Ohlendorf, W.; Poschenrieder, W.; Rapp, H.; Rau, F.; Renner, H.; Riedler, H.; Ringler, H.; Sardei, F.; Schüller, P. G.; Schwörer, K.; Siller, G.; Söldner, F.; Steuer, K.-H.; Thumm, M.; Tutter, M.; Vollmer, O.; Weller, A.; Wilhelm, R.; Wobig, H.; Würsching, E.; Zippe, M.

    1986-05-01

    Both in ohmically and beam-heated L-mode discharges of ASDEX, the electron-temperature (Te) profile shape can be varied over a wide range by the choice of the safety factor qa. The power-deposition profile, on the contrary, has no effect on the Te-profile shape. In current-free WVII-A stellarator plasmas, no such invariance property is found. An independent constraint seems to fix the current distribution j(r) of the tokamak, which defines the conditions of electron heat transport.

  14. Vacuum Performance Improvement of the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    LIUDequan; CAOZeng; CUIChenghe; RANHong; LUPing; XIEYanfeng; XUZenghua

    2003-01-01

    There are many inner items in vacuum vessel, and hundreds of seals on the vessel borts in the HL-2A tokamak, so it is difficult to get excellent vacuum. During the engineering test in 2002, the results of quadrupole mass spectrometer showed thatthe partial pressures of oxygen and H2O were higher than normal value, some leaks on theupper MP3 coil jackets and port seals had been detected. In order to carry out good discharge experiment, the good vacuum and wall condition must be maintained.

  15. A first wall material probe manipulator for the 'TEXTOR' tokamak

    International Nuclear Information System (INIS)

    Textor is a technology oriented tokamak of Euratom at the Kernforschungsanlage Juelich (KFA). Switzerland participates in its experimental program within the framework of the IEA agreement on Plasma Wall Interaction. A major task of EIR consists in the layout, construction and fabrication of a manipulator for the remote handling of up to 240 specimen candidate first wall materials. This operation has to be done without breaking the ultra high vacuum (UHV) and with wall temperatures up to 3000C. A great number of preexperiments involving different materials had to be carried out; the understanding of the tribology in ultra high vacuum could be improved. (Auth.)

  16. Unified Description of Tokamak Ideal MHD Instabilities (Ⅰ)

    Institute of Scientific and Technical Information of China (English)

    石秉仁

    2002-01-01

    By using a coordinate system associated with magnetic surfaces, a unified eigen mode equation for describing the tokamak ideal MHD instabilities is derived in the shear-Alfven approximation. Based on this equation having a general operator form, the eigen-mode equation governing the large-scale perturbation (such as the kink mode, the low-n ballooning mode and the Alfven mode) and small-scale perturbation (such as the high-n ballooning mode, the local mode)can be further deduced. In the first part of the present study, the small-scale perturbation is discussed in detail.

  17. Unified Description of Tokamak Ideal MHD Instabilities(I)

    Institute of Scientific and Technical Information of China (English)

    石秉仁

    2002-01-01

    By using a coordinate system associated with magnetic surfaces,a unified eigenmode equation for describing the tokamak ideal MHD instabilities is derived in the shear-Alfven approximation.Based on this equation having a general operator form,the eigen-mode equation governing the large-scale perturbation (such as the kink mode,the low-n ballooning mode and the Alfven mode) and small-scale perturbation(such as the high-n ballooning mode,the local mode) can be further deduced.In the first part of the present study,the small-scale perturbation is discussed in detail.

  18. Tokamak-7 cryogenic system and its test data

    International Nuclear Information System (INIS)

    The composition, structure and purpose of the Tokamak-7 device cryogenic system are described. The cryogenic system comprises the helium system and nitrogen cryogenic system. Heat inflows at different temperature levels to cryogenic objects under various modes of the device operation are estimated. The results of the cryogenic system autonomous tests under cooling, cryostating and heating of the T-7 device cryogenic objects are presented. The three year experience of cryogenic system operation for the circulation superconducting magnetic system is accumulated and analysed. Important results necessary for designing cryogenic complexes of newly developed large thremonuclear devices are obtained

  19. Eikonal waves, caustics and mode conversion in tokamak plasmas

    Science.gov (United States)

    Jaun, A.; Tracy, E. R.; Kaufman, A. N.

    2007-01-01

    Ray optics is used to model the propagation of short electromagnetic plasma waves in toroidal geometry. The new RAYCON code evolves each ray independently in phase space, together with its amplitude, phase and focusing tensor to describe the transport of power along the ray. Particular emphasis is laid on caustics and mode conversion layers, where a linear phenomenon splits a single incoming ray into two. The complete mode conversion algorithm is described and tested for the first time, using the two space dimensions that are relevant in a tokamak. Applications are shown, using a cold plasma model to account for mode conversion at the ion-hybrid resonance in the Joint European Torus.

  20. Transport threshold model of subsonic neoclassical tearing modes in tokamaks

    International Nuclear Information System (INIS)

    A transport threshold model of subsonic neoclassical tearing modes (NTMs) in tokamaks is developed. The basic procedure is to include the velocity-dependent term into the single-fluid heat-conductivity equation and to complement this equation with the single-fluid parallel plasma motion equation. These equations permit the determination of the perturbed plasma temperature and the bootstrap current drive of NTMs, for both strong and weak perpendicular heat transport, which is the precondition for developing the above model. It is shown that the subsonic NTMs transport threshold model can be more realistic than the standard transport model of NTMs suggested by Fitzpatrick [Phys. Plasmas 2, 825 (1995)

  1. Experimental setup for tungsten transport studies at the NSTX tokamak

    International Nuclear Information System (INIS)

    Tungsten particles have been introduced into the National Spherical Torus Experiment (NSTX) in Princeton with the purpose to investigate the effects of tungsten injection on subsequent plasma discharges. An experimental setup for the study of tungsten particle transport is described where the particles are introduced into the tokamak using a modified particle dropper, otherwise used for lithium-powder injection. An initial test employing a grazing-incidence extreme ultraviolet spectrometer demonstrates that the tungsten-transport setup could serve to infer particle transport from the edge to the hot central plasmas of NSTX.

  2. A design retrospective of the DIII-D tokamak

    Science.gov (United States)

    Luxon, J. L.

    2002-05-01

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

  3. Controlling tokamak geometry with three-dimensional magnetic perturbations

    International Nuclear Information System (INIS)

    It is shown that small externally applied magnetic perturbations can significantly alter important geometric properties of magnetic flux surfaces in tokamaks. Through 3D shaping, experimentally relevant perturbation levels are large enough to influence turbulent transport and MHD stability in the pedestal region. It is shown that the dominant pitch-resonant flux surface deformations are primarily induced by non-resonant 3D fields, particularly in the presence of significant axisymmetric shaping. The spectral content of the applied 3D field can be used to control these effects

  4. First Results from the Lithium Tokamak eXperiment

    International Nuclear Information System (INIS)

    Full text: The Lithium Tokamak eXperiment (LTX) is a newly commissioned, modest-scale spherical tokamak with R =0.4 m, a =0.26 m, and elongation of 1.5. Design targets are a toroidal field of 3.2 kG, plasma current up to 400 kA, and a discharge duration of order 100 msec. LTX is the first tokamak designed to investigate modifications to equilibrium and transport when global recycling is reduced to 10 - 20 %. To reduce recycling, LTX is fitted with a 1 cm thick heated (300 deg. C) copper shell, conformal to the last closed flux surface, over 85% of the plasma surface area. The plasma-facing surface of the shell will be evaporatively coated with a thin (< 100 micron) layer of molten lithium, retained by surface tension. The shell is replaceable, and a second version has been constructed, which was plasma-sprayed with 100 - 200 microns of molybdenum to form a high-Z substrate for subsequent coating with lithium. After the installation of the second shell (in 2011), a high temperature (500 - 600 deg. C) operating phase for LTX is planned. LTX is the first tokamak designed to operate with a full hot high-Z wall, near the projected operating temperature for reactor PFCs. The engineering design and construction of the hot high-Z shell, as well as the vessel and diagnostics to tolerate both lithium and 500 deg. C internal components, will be discussed. LTX will employ short-pulse fueling with a new hydrogen molecular cluster injector, to transiently eliminate edge gas (between puffs). This fueling system will be briefly discussed. Diagnostics include single-pulse multipoint Thomson scattering, Lyman alpha arrays, microwave interferometers, spectrometers, and an edge Langmuir probe. LTX is now progressing through the shakedown phase, and first operation with a liquid lithium film wall is scheduled for Spring 2010. Later in 2010, a new diagnostic (Digital Holography) for core density variations will be tested on LTX. In 2011 a 5 A, 20 kV, 1 second hydrogen neutral beam

  5. Real-Time Software for the Compass Tokamak Plasma Control

    Energy Technology Data Exchange (ETDEWEB)

    Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Prague (Costa Rica)

    2009-07-01

    This poster presents the flexible and high-performance real time system that guarantees the desired time cycles for plasma control on the COMPASS tokamak: 500 {mu}s for toroidal field, current, equilibrium and shaping; 50 {mu}s for fast control of the equilibrium and vertical instability. This system was developed on top of a high-performance processor and a software framework (MARTe) tailored for real-time. The preliminary measurements indicate that the time constraints will be met on the final solution. The system allows the making of modifications in the future to improve software components. (A.C.)

  6. First-wall sputtering effects in Cat-D tokamaks

    International Nuclear Information System (INIS)

    Cat-D tokamaks offer the important advantages of elimination of the need to breed tritium and less frequent first-wall replacement due to reduced neutron damage effects. To capitalize on these features, however, the erosion of the low-Z first-wall coating must be addressed. As shown here, this appears to be feasible if a highly efficient (beta > 0.9) mechanical limiter system can be developed. Still, due to the many uncertainties involved, the simultaneous development of in situ techniques for redeposition coatings appears desirable

  7. Effect of Recycling in the HL-1M Tokamak

    Institute of Scientific and Technical Information of China (English)

    郑永真

    2004-01-01

    Tokamak plasma discharge disruption at high density is investigated. The instability analysis on model indicates that the disruption is resulted from the energy loss arising from hydrogen recycling on the edge of the plasma. This energy loss could lead to a contraction of the current channel and the production of a disruptively unstable configuration. Using a simple model we shall investigate the implications of recycling for disruptions. The critical high-density n ≤ 1.6 × 10 20 m-3 is reached in LH-1M.

  8. The tokamak density limit: A thermo-resistive disruption mechanism

    Science.gov (United States)

    Gates, David

    2015-11-01

    A magnetic island growth mechanism based on radiative cooling of the internal island flux surfaces is shown to produce the correct physical scaling to explain one of the long standing mysteries of tokamak physics - the empirical Greenwald density limit. In this presentation we will review the phenomenology of the density limit and the correlation between the Greenwald limit and the onset threshold for radiation-driven tearing modes. The behavior of magnetic islands with a 3D electron temperature distribution which is consistent with a large ratio of radial to parallel heat conductivity - and a corresponding 3D resistivity profile - is examined for islands with near-zero net heating in the island interior. The effect of varying impurity mix on the local island onset threshold is consistent with the established experimental scalings for tokamaks at the density limit. A simple analytic theory is developed which reveals the effect of heating and cooling in the island interior as well as the effect of island asymmetry. It is shown that a new term accounting for the thermal effects of island asymmetry is a crucial addition to the Modified Rutherford Equation. The resultant model exhibits a robust onset of a rapidly growing tearing mode - consistent with the disruption mechanism observed at the density limit in tokamaks. Additionally, a fully non-linear 3D cylindrical calculation is performed that simulates the effect of net island heating / cooling by raising / suppressing the temperature in the core of the island. In both the analytic theory and the numerical simulation a sudden threshold for explosive growth is found to be due to the interaction between three distinct thermal non-linearities, which affect the island resistivity, thereby modifying the growth dynamics. Expanding on the model presented, we speculate that the mechanism described may be applicable to a much wider range of tokamak disruptions than just those near the Greenwald limit. This work is supported

  9. Spectral measurements of runway electrons in the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kudyakov, Timur

    2009-07-22

    The generation of multi-MeV runaway electrons is a well known effect related to the plasma disruptions in tokamaks. The runaway electrons can substantially reduce the lifetime of the future tokamak ITER. In this thesis physical properties of runaway electrons and their possible negative effects on ITER have been studied in the TEXTOR tokamak. A new diagnostic, a scanning probe, has been developed to provide direct measurements of the absolute number of runaway electrons coming from the plasma, its energy distribution and the related energy load in the material during low density (runaway) discharges and during disruptions. The basic elements of the probe are YSO crystals which transform the energy of runaway electrons into visible light which is guided via optical fibres to photomultipliers. In order to obtain the energy distribution of runaways, the crystals are covered with layers of stainless steel (or tungsten in two earlier test versions) of different thicknesses. The final probe design has 9 crystals and can temporally and spectrally resolve electrons with energies between 4 MeV and 30 MeV. The probe is tested and absolutely calibrated at the linear electron accelerator ELBE in Rossendorf. The measurements are in good agreement with Monte Carlo simulations using the Geant4 code. The runaway transport in the presence of the internal and externally applied magnetic perturbations has been studied. The diffusion coefficient and the value of the magnetic fluctuation for runaways were derived as a function of B{sub t}. It was found that an increase of runaway losses from the plasma with the decreasing toroidal magnetic field is accompanied with a growth of the magnetic fluctuation in the plasma. The magnetic shielding picture could be confirmed which predicts that the runaway loss occurs predominantly for low energy runaways (few MeV) and considerably less for the high energy ones. In the case of the externally applied magnetic perturbations by means of the dynamic

  10. Calculation of plasma position and shape in KT-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Doo Hee; Chung Kyu Sun [Hanyang Univ., Seoul (Korea, Republic of); Oh, Byung Hoon; Jeong, Seung Ho; Hong, Bong Guen; Lee, Kwang Won [KAERI, Taejon (Korea, Republic of)

    2000-05-01

    For the first time, the full-time scale variations of plasma position and shape from the outmost poloidal isoflux surfaces during the buildup, flat-top, and decay period of plasma current are calculated from the measurement of magnetic fields at the plasma-boundary surface in a KT-1 Tokamak (major radius of 27 cm and minor radius of 4.2 cm). Three kinds of magnetic probes arranged poloidally in 30 degrees outside the torus are used in determining the plasma boundary. Without the information of internal plasma current profiles, the calculations are only performed by a linear combination of measured fields.

  11. Photometric Calibration of the SPRED at the FTU Tokamak

    International Nuclear Information System (INIS)

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

  12. Continuum Kinetic Modeling of the Tokamak Plasma Edge

    Science.gov (United States)

    Dorf, Mikhail

    2015-11-01

    The problem of edge plasma transport provides substantial challenges for analytical or numerical analysis due to (a) complex magnetic geometry including both open and closed magnetic field lines B, (b) steep radial gradients comparable to ion drift-orbit excursions, and (c) a variation in the collision mean-free path along B from long to short compared to the magnetic connection length. Here, the first 4D continuum drift-kinetic transport simulations that span the magnetic separatrix of a tokamak are presented, motivated in part by the success of continuum kinetic codes for core physics and in part by the potential for high accuracy. The calculations include fully-nonlinear Fokker-Plank collisions and electrostatic potential variations. The problem of intrinsic toroidal rotation driven by ion orbit loss is addressed in detail. The code, COGENT, developed by the Edge Simulation Laboratory collaboration, is distinguished by a fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex magnetic X-point divertor geometry with high accuracy. Previously, successful performance of high-order algorithms has been demonstrated in a simpler closed magnetic-flux-surface geometry for the problems of neoclassical transport and collisionless relaxation of geodesic acoustic modes in a tokamak pedestal, including the effects of a strong radial electric field under H-mode conditions. Work performed for USDOE, at LLNL under contract DE-AC52-07NA27344.

  13. In-vessel remote maintenance of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

  14. Current drive and profile control in low aspect ratio tokamaks

    International Nuclear Information System (INIS)

    The key to the theoretically predicted high performance of a low aspect ratio tokamak (LAT) is its ability to operate at very large plasma current*Ip. The plasma current at low aspect ratios follows the approximate formula: Ip ∼ (5a2Bt/Rqψ) [(1 + κ2)/2] [A/(A - 1)] where A quadruple-bond R/a which was derived from equilibrium studies. For constant qψ and Bt, Ip can increase by an order of magnitude over the case of tokamaks with A approx-gt 2.5. The large current results in a significantly enhanced βt (quadruple-bond βNIp/aBt) possibly of order unity. It also compensates for the reduction in A to maintain the same confinement performance assuming the confinement time τ follows the generic form ∼ HIpP-1/2R3/2κ1/2. The initiation and maintenance of such a large current is therefore a key issue for LATs

  15. The vacuum vessel thermal shield of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, B.J. E-mail: bjyoon@kaeri.re.kr; In, S.R.; Cho, S.Y

    2003-09-01

    The Korea superconducting tokamak advanced research (KSTAR) tokamak has an all-superconductor magnet system and needs a thermal shield to cut off thermal radiation from the components of room temperature. The vacuum vessel thermal shield (VVTS) cooled to 70 K is placed in the narrow gap between the 5 K TF magnets and the 300 K vacuum vessel (VV). The VVTS is designed to be divided into 16 assembly modules of 22.5 deg. sector, each unit has an electrical insulation along the center line in the toroidal direction and four insulations in the poloidal direction to reduce eddy currents induced during plasma operations. All connections are bolted. The VVTS becomes consequently a rigid torus composed of 64 electrically insulated pieces. A key point of designing the VVTS is that supports of the VVTS are to be flexible enough to allow thermal constriction during cooling down to 70 K as well as sufficiently strong to withstand electromagnetic (EM) forces exerted on the VVTS during plasma disruptions. Leaf spring type supports devised to satisfy these requirements are to be installed along the mid plane of the VVTS. The cryopanel of the VVTS is of quilted plate type whose total thickness is 12 mm, cooled by 60 K, 20 bar GHe.

  16. Nuclear shielding of openings in ITER Tokamak building

    Energy Technology Data Exchange (ETDEWEB)

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Arumugam, A.P.; Beaudoin, V.; Beltran, D.; Benchikhoune, M.; Berruyer, F.; Cortes, P.; Gandini, F. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghirelli, N. [ASSYSTEM E.O.S, ZAC Saint Martin, 23, rue Benjamin Franklin, 84120 Pertuis (France); Gray, A.; Hurzlmeier, H.; Le Page, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Lentini, G.; Loughlin, M.; Mita, Y.; Patisson, L.; Rigoni, G.; Rathi, D.; Song, I. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.

  17. The timing system on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Wei [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhang, Ming, E-mail: zhangming@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhuang, Ge; Ding, Tonghai; Huang, Fuqiang; Shan, Lingjie [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-01-15

    Highlights: •The timing system achieved tree structured timing network with only one type of timing module. •This system is integrated into J-TEXT COADC which is an EPICS based control system. •This system handles multiple timing sequences and events. •This system has been deployed on J-TEXT and working properly in daily experiments. -- Abstract: This paper describes the timing system designed to control the operation time-sequence and to generate clocks for various sub-systems on J-TEXT tokamak. The J-TEXT timing system is organized as a distributed system which is connected by a tree-structured optical fiber network. It can generate delayed triggers and gate signals (0 μs–4000 s), while providing reference clocks for other sub-systems. Besides, it provides event handling and timestamping functions. It is integrated into the J-TEXT Control, Data Access and Communication (J-TEXT CODAC) system, and it can be monitored and configured by Experimental Physics and Industrial Control System (EPICS). The configuration of this system including tree-structured network is managed in XML files by dedicated management software. This system has already been deployed on J-TEXT tokamak and it is serving J-TEXT in daily experiments.

  18. Power balance in the divertor-tokamak DIVA

    International Nuclear Information System (INIS)

    Power balances of Ohmically and radio-frequency (RF) heated plasmas including a boundary (scrape-off layer) plasma are investigated in the divertor-tokamak DIVA. First, methods of measurement of the boundary plasma are described. These are applied to the divertor plasma in the case of Ohmic heating. The results clarify characteristics of the boundary plasma of a conventional tokamak. The scaling law for the boundary plasma is derived in consideration of the power balance including the boundary plasma. Heat flux to material surfaces is investigated in detail; the relationship between heat flux, particle flux and electron velocity distribution is clarified. Gross power balance is investigated by measurements of total heat flux to the wall and total radiation loss including charge-exchange loss. These results provide experimental evidence for the above scaling law. Finally, power balance during the Ion-Cyclotron Range of Frequency (ICRF) heating is described. Optimum heating conditions of the ICRF heating in the two-ion hybrid regime are surveyed. For the optimum heating conditions, gross power balance including the boundary plasma is considered, in which the heating efficiency is derived. Radial profile of the RF-heating power, the ratio of the heating power to each species and the transport of RF-heated ions are clarified in the power balance. (author)

  19. Investigation of magnetic modes in the ASDEX tokamak

    International Nuclear Information System (INIS)

    Properties of MHD-modes in the ASDEX Tokamak have been investigated by application and further development of the MIRNOV-diagnostics, i.e. measurement of magnetic field fluctuations. In addition to evaluation methods supported by models, also a model-independent statistical data analysis makes sense. The very important physics of mode locking, i.e. the slowing-down of rotating modes is examined. An elaborated theoretical model allows an interpretation of experimental results. Especially interesting is the loss of the angular momentum of rotating plasmas by mode locking. Experiments for mode stabilisation and prevention of electric current breakdown are discussed. Additional MHD-processes under different plasma conditions are treated on the fundament of the devloped model ideas. The author shows that the main tokamak plasma is described very well by one-dimensional models with cylindrical geometry, while the boundary zone of the plasma demands a more complex analysis. In the appendix a concept for the investigation of the MHD-activity in ASDEX-Upgrade is discussed. (AH)

  20. Overview of the EUROfusion Medium Size Tokamak program

    Science.gov (United States)

    Martin, Piero; Beurskens, Marc; Coda, Stefano; Eich, Thomas; Meyer, Hendrik; the EUROfusion MST1 Team

    2015-11-01

    As a result of the new organization of the European fusion programme, now under the umbrella of the EUROfusion Consortium, the MST (Medium Size Tokamaks) task force is in charge of executing the European science programme in the ASDEX Upgrade, TCV and MAST-U tokamaks. This paper will present an overview of the main results obtained in the 2014 campaign-where only ASDEX upgrade was operating-and the preliminary achievements of the recently started 2015/16 campaign, where also TCV will contribute. The main subjects of the experimental campaigns are (i) the development of scenarios relevant for the ITER Q=10 goal, in an all metal wall device (ii) the understanding of ELM mitigation/suppression with pellets and resonant magnetic perturbations, and in particular the effect of density versus collisionality, (iii) the understanding and optimization of methods for disruption mitigation or avoidance and runaway electrons control and (iv) the exploration of ITER and DEMO relevant scenarios with high normalized separatrix power flux, Psep / R , (Psep is the power through the separatrix, R the major radius) and tolerable target heat loads. The overview of the future programs in MST will be given. http://www.euro-fusionscipub.org/mst1

  1. Anomalous energy transport in hot plasmas: solar corona and Tokamak

    International Nuclear Information System (INIS)

    Anomalous energy transport is studied in two hot plasmas and appears to be associated with a heating of the solar corona and with a plasma deconfining process in tokamaks. The magnetic structure is shown to play a fundamental role in this phenomenon through small scale instabilities which are modelized by means of a nonlinear dynamical system: the Beasts' Model. Four behavior classes are found for this system, which are automatically classified in the parameter space thanks to a neural network. We use a compilation of experimental results relative to the solar corona to discuss current-based heating processes. We find that a simple Joule effect cannot provide the required heating rates, and therefore propose a dimensional model involving a resistive reconnective instability which leads to an efficient and discontinuous heating mechanism. Results are in good agreement with the observations. We give an analytical expression for a diffusion coefficient in tokamaks when magnetic turbulence is perturbing the topology, which we validate thanks to the standard mapping. A realistic version of the Beasts' Model allows to test a candidate to anomalous transport: the thermal filamentation instability

  2. Neoclassical Physics for Current Drive in Tokamak Plasmas

    International Nuclear Information System (INIS)

    The Lie transform formalism is applied to charged particle dynamics in tokamak magnetic topologies, in order to build a Fokker-Planck type operator for Coulomb collisions usable for current drive. This approach makes it possible to reduce the problem to three dimensions (two in velocity space, one in real space) while keeping the wealth of phase-space cross-term coupling effects resulting from conservation of the toroidal canonical momentum (axisymmetry). This kinetic approach makes it possible to describe physical phenomena related to the presence of strong pressure gradients in plasmas of an unspecified form, like the bootstrap current which role will be paramount for the future ITER machine. The choice of coordinates and the method used are particularly adapted to the numerical resolution of the drift kinetic equation making it possible to calculate the particle distributions, which may present a strong variation with respect to the Maxwellian under the effect of an electric field (static or produced by a radio-frequency wave). This work, mainly dedicated to plasma physics of tokamaks, was extended to those of space plasmas with a magnetic dipole configuration. (author)

  3. Recent lithium experiments in tokamak T-11M

    Energy Technology Data Exchange (ETDEWEB)

    Mirnov, S.V., E-mail: mirnov@triniti.ru [SSC PF TRINITI, 142 190 Troitsk, Moscow (Russian Federation); Belov, A.M.; Djigailo, N.T. [SSC PF TRINITI, 142 190 Troitsk, Moscow (Russian Federation); Kostina, A.N. [NRNU MEPhI, Kashirskoye sh. 31, Moscow 115 409 (Russian Federation); Lazarev, V.B. [SSC PF TRINITI, 142 190 Troitsk, Moscow (Russian Federation); Lyublinski, I.E. [JSC “Red Star”, Elektrolitnyj pr. 1A, Moscow 113 230 (Russian Federation); Nesterenko, V.M. [SSC PF TRINITI, 142 190 Troitsk, Moscow (Russian Federation); Vertkov, A.V. [JSC “Red Star”, Elektrolitnyj pr. 1A, Moscow 113 230 (Russian Federation)

    2013-07-15

    The T-11M lithium program is focused to find solutions for technological issues of a steady-state tokamak. A recent lithium activity in T-11M was focused on three directions: investigation of different Li-limiters, investigations of long-term effects of lithium behavior in hydrogen plasmas and development of a new kind of CPS limiters. The new so-called R limiter has been used as Li collector for investigation of lithium fluxes near the plasma boundary and chamber walls. It was shown, that the Li circulation near the limiter exceeds up to four times the primary lithium flux from limiters. 90% of primary flux is collected by lateral sides of limiter and only 10% arrives at the chamber wall. The deuterium glow discharge was used to test the long-term lithium degradation under the deuterium bombardment. It was shown that such degradation is small. That means that the Li-limiter can be used as Li-emitter in steady-state tokamak operation. A new vertical lithium limiter was successfully tested in T-11M.

  4. Charge exchange recombination spectroscopy on the T-10 tokamak.

    Science.gov (United States)

    Klyuchnikov, L A; Krupin, V A; Nurgaliev, M R; Korobov, K V; Nemets, A R; Dnestrovskij, A Yu; Tugarinov, S N; Serov, S V; Naumenko, N N

    2016-05-01

    The charge exchange recombination spectroscopy (CXRS) diagnostics on the T-10 tokamak is described. The system is based on a diagnostic neutral beam and includes three high etendue spectrometers designed for the ITER edge CXRS system. A combined two-channel spectrometer is developed for simultaneous measurements of two beam-induced spectral lines using the same lines of sight. A basic element of the combined spectrometer is a transmitting holographic grating designed for the narrow spectral region 5291 ± 100 Å. The whole CXRS system provides simultaneous measurements of two CXRS impurity spectra and Hα beam line. Ion temperature measurements are routinely provided using the C(6+) CXRS spectral line 5291 Å. Simultaneous measurements of carbon densities and one more impurity (oxygen, helium, lithium etc.) are carried out. Two light collecting systems with 9 lines of sight in each system are used in the diagnostics. Spatial resolution is up to 2.5 cm and temporal resolution of 1 ms is defined by the diagnostic neutral beam diameter and pulse duration, respectively. Experimental results are shown to demonstrate a wide range of the CXRS diagnostic capabilities on T-10 for investigation of impurity transport processes in tokamak plasma. Developed diagnostics provides necessary experimental data for studying of plasma electric fields, heat and particle transport processes, and for investigation of geodesic acoustic modes. PMID:27250422

  5. An emerging understanding of H-mode discharges in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Groebner, R.J.

    1992-12-01

    A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the {upsilon}{sub E}{sup {yields}} = (E {times} B)/B{sup 2} flow velocity. These results are qualitatively consistent with theories which predict suppression of fluctuations by shear or curvature in {upsilon}E. The required {upsilon}E flow is generated very rapidly when the magnitude of the heating power or of an externally imposed radial current exceed threshold values and several theoretical models have been developed to explain the observed changes in the {upsilon}E flow. After the transition occurs, the altered boundary conditions enable the development of improved confinement in the plasma interior on a confinement time scale. The resulting H-mode discharge has typically twice the confinement of L-mode discharges and regimes of further improved confinement have been obtained in some H-mode scenarios.

  6. Conceptual design of a commercial tokamak reactor using resistive magnets

    International Nuclear Information System (INIS)

    The potential of resistive magnet tokamaks as commercial electricity producing power plants is investigated. Parametric studies indicate that attractive design space exists for these reactors at relatively low field (2.5 to 4.5 T), moderate wall loading (3 to 4 MW/m2) and medium to large net electric outputs (>600 MW/sub e/). High toroidal beta (20 to 25%) possible in the second regime of plasma stability may provide advantages of reduced recirculating power and plasma current but moderate beta reactors (6 to 10%) remain attractive. A conceptual design for the Resistive magnet Commercial Tokamak Reactor (RCTR) is presented. The layout of the nuclear island is driven by compatibility requirements of the demounting capability with structural and blanket design considerations. The nuclear island is fully demountable with access to all components within the toroidal field coils possible via simple vertical lifts. The blanket system, segmented for vertical removal, uses a self-cooled liquid lithium breeder/coolant with vanadium structure and an HT-9 reflector. The first wall is also lithium cooled with a vanadium structure but is constructed in a single, pre-tested unit for assembly and periodic replacement. Ohmic and equilibrium field-coils are located within the bore of the toroidal field coil for improved performance

  7. Analysis of fast ion induced instabilities in tokamak plasmas

    CERN Document Server

    Horváth, László

    2015-01-01

    In magnetic confinement fusion devices like tokamaks, it is crucial to confine the high energy fusion-born helium nuclei ($\\alpha$-particles) to maintain the energy equilibrium of the plasma. However, energetic ions can excite various instabilities which can lead to their enhanced radial transport. Consequently, these instabilities may degrade the heating efficiency and they can also cause harmful power loads on the plasma-facing components of the device. Therefore, the understanding of these modes is a key issue regarding future burning plasma experiments. One of the main open questions concerning energetic particle (EP) driven instabilities is the non-linear evolution of the mode structure. In this thesis, I present my results on the investigation of $\\beta$-induced Alfv\\'{e}n eigenmodes (BAEs) and EP-driven geodesic acoustic modes (EGAMs) observed in the ramp-up phase of off-axis NBI heated plasmas in the ASDEX Upgrade tokamak. These modes were well visible on several line-of-sights (LOSs) of the soft X-ra...

  8. Line-integrated emissivity in tokamak plasma radiation diagnostics

    International Nuclear Information System (INIS)

    In tokamak plasma radiation diagnostics, an emissivity profile derived from line-integrated radiation measurement data with Abel inversion or tomographic analysis is averaged over a poloidal cross-section of the plasma chord. The width of this cross-section restricts the spatial resolution of the given diagnostic system. In the cases where the condition for infinitesimal-detector approximation is not satisfied, this width appears to be markedly greater than that defined by the angle viewed from the detector center to the poloidal dimension of an aperture, and the contributions to the detected signal from different points on the given cross-section of the plasma chord are not the same, so the average of the emissivity over the cross-section is a weighted one. In fact, many radiation diagnostic systems on tokamaks do not satisfy the infinitesimal-detector approximation condition. On the basis of the derived explicit formula for the line-integrated emissivity, the relevant questions are discussed in detail

  9. Realtime capable first principle based modelling of tokamak turbulent transport

    Science.gov (United States)

    Citrin, Jonathan; Breton, Sarah; Felici, Federico; Imbeaux, Frederic; Redondo, Juan; Aniel, Thierry; Artaud, Jean-Francois; Baiocchi, Benedetta; Bourdelle, Clarisse; Camenen, Yann; Garcia, Jeronimo

    2015-11-01

    Transport in the tokamak core is dominated by turbulence driven by plasma microinstabilities. When calculating turbulent fluxes, maintaining both a first-principle-based model and computational tractability is a strong constraint. We present a pathway to circumvent this constraint by emulating quasilinear gyrokinetic transport code output through a nonlinear regression using multilayer perceptron neural networks. This recovers the original code output, while accelerating the computing time by five orders of magnitude, allowing realtime applications. A proof-of-principle is presented based on the QuaLiKiz quasilinear transport model, using a training set of five input dimensions, relevant for ITG turbulence. The model is implemented in the RAPTOR real-time capable tokamak simulator, and simulates a 300s ITER discharge in 10s. Progress in generalizing the emulation to include 12 input dimensions is presented. This opens up new possibilities for interpretation of present-day experiments, scenario preparation and open-loop optimization, realtime controller design, realtime discharge supervision, and closed-loop trajectory optimization.

  10. A Research Program of Spherical Tokamak in China

    Institute of Scientific and Technical Information of China (English)

    何也熙

    2002-01-01

    The mission of this program is to explore the spherical torus plasma with a SUNIST spherical tokamak. Main experiments in the start phase will be involved with breakdown and plasma current set-up with a mode of saving volt-second and without ohmic heating system, equilibrium and instability, current driving, heating and profile modification. The SUNIST is a university-scale conceptual spherical tokamak, with R = 0.3 m, A 1.3, Ip ~ 50 kA, BT < 0.15 T, and PRF = 100 kW. The only peculiarity of SUNIST is that there is a toroidal insulating break along the outer wall of vacuum vessel. The expected that advantages of this arrangement are helpful not only for saving flux swing, but also for having a deep understanding of what will influence the discharge startup and globe performances of plasma under different conditions of strong vessel eddy and ECR power assistance. Of course, the vessel structure of cross seal will be at a great risk of controlling vacuum quality, although we have achieved positive results on simulation test and vacuum vessel test.

  11. The GBS code for tokamak scrape-off layer simulations

    Science.gov (United States)

    Halpern, F. D.; Ricci, P.; Jolliet, S.; Loizu, J.; Morales, J.; Mosetto, A.; Musil, F.; Riva, F.; Tran, T. M.; Wersal, C.

    2016-06-01

    We describe a new version of GBS, a 3D global, flux-driven plasma turbulence code to simulate the turbulent dynamics in the tokamak scrape-off layer (SOL), superseding the code presented by Ricci et al. (2012) [14]. The present work is driven by the objective of studying SOL turbulent dynamics in medium size tokamaks and beyond with a high-fidelity physics model. We emphasize an intertwining framework of improved physics models and the computational improvements that allow them. The model extensions include neutral atom physics, finite ion temperature, the addition of a closed field line region, and a non-Boussinesq treatment of the polarization drift. GBS has been completely refactored with the introduction of a 3-D Cartesian communicator and a scalable parallel multigrid solver. We report dramatically enhanced parallel scalability, with the possibility of treating electromagnetic fluctuations very efficiently. The method of manufactured solutions as a verification process has been carried out for this new code version, demonstrating the correct implementation of the physical model.

  12. Real-time software for the COMPASS tokamak plasma control

    Energy Technology Data Exchange (ETDEWEB)

    Valcarcel, D.F., E-mail: danielv@ipfn.ist.utl.p [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic)

    2010-07-15

    The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 {mu}s. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.

  13. Electric potential cells at the diverted tokamak separatrix

    International Nuclear Information System (INIS)

    Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (pe) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼100 V) drives ExB circulation ('potential cell') of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arises from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the ∇B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. We speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode. (author)

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

    International Nuclear Information System (INIS)

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

  15. Second regime tokamak operation at large aspect ratio

    International Nuclear Information System (INIS)

    The equilibrium, stability, and transport properties of large aspect ratio tokamaks operating at the second stability regime are described theoretically using numerical and analytical techniques. It has been shown that, at large aspect ratio, significant current profile control is possible with relatively modest amounts of neutral beam current drive, and the power needed to access and maintain the second regime operation is calculated to be about 3 MW using the results of an integrated 1.5D transport and stability code. An example second regime experiment has been described and the results are presented of extensive calculations illustrating several possible operating scenarios, external and internal model stability boundaries, and the experimental features needed to evaluate and test the high beta tokamak theories. The theory which describes the stabilizing effect of energetic particles during high beta operation was extended to finite aspect ratio. A key technical problem for application of this technique appears to be caused by ripple transport. Plasma rotation effects are found to be generally destabilizing and several other schemes for improved access to the second stability regime are discussed including ponderomotive stabilization of the plasma edge region and active feedback control

  16. Finite pressure effects on the tokamak sawtooth crash

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Yasutaro

    1998-07-01

    The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed.

  17. Finite pressure effects on the tokamak sawtooth crash

    International Nuclear Information System (INIS)

    The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed

  18. Nuclear shielding of openings in ITER Tokamak building

    International Nuclear Information System (INIS)

    Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different

  19. Observation of compressional Alfven eigenmodes (CAE) in a conventional tokamak

    International Nuclear Information System (INIS)

    Fast-ion instabilities with frequencies somewhat below the ion cyclotron frequency occur frequently in spherical tokamaks such as the National Spherical Torus Experiment (NSTX). NSTX and the DIII-D tokamak are nearly ideal for fast-ion similarity experiments, having similar neutral beams, fast-ion to Alfven speed vf/vA, fast-ion pressure, and shape of the plasma but with a factor of two difference in major radius. When DIII-D is operated at low field (0.6 T), compressional Alfven eigenmode (CAE) instabilities appear that closely resemble the NSTX instabilities. In particular, the mode frequencies, polarization and beam-energy threshold are nearly identical to NSTX. CAE in high-field discharges and emission at cyclotron harmonics are also observed. As on NSTX, the basic stability properties are consistent with the idea that the instability is driven by anisotropy in the fast-ion velocity distribution and is damped predominantly by Landau damping of electrons. The results suggest that these modes might be excited in ITER

  20. The tokamak density limit: A thermo-resistive disruption mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Gates, D. A.; Brennan, D. P.; Delgado-Aparicio, L.; White, R. B. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543 (United States)

    2015-06-15

    The behavior of magnetic islands with 3D electron temperature and the corresponding 3D resistivity effects on growth are examined for islands with near-zero net heating in the island interior. We refer to the resulting class of non-linearities as thermo-resistive effects. In particular, the effects of varying impurity mix on the previously proposed local island onset threshold [Gates and Delgado-Aparicio, Phys. Rev. Lett. 108, 165004 (2012)] are examined and shown to be consistent with the well established experimental scalings for tokamaks at the density limit. A surprisingly simple semi-analytic theory is developed which imposes the effects of heating/cooling in the island interior as well as the effects of island geometry. For the class of current profiles considered, it is found that a new term that accounts for the thermal effects of island asymmetry is required in the modified Rutherford equation. The resultant model is shown to exhibit a robust onset of a rapidly growing tearing mode—consistent with the disruption mechanism observed at the density limit in tokamaks. A fully non-linear 3D cylindrical calculation is performed that simulates the effect of net island heating/cooling by raising/suppressing the temperature in the core of the island. In both the analytic theory and the numerical simulation, the sudden threshold for rapid growth is found to be due to an interaction between three distinct thermal non-linearities which affect the island resistivity, thereby modifying the growth dynamics.

  1. Kinetic analysis of MHD ballooning modes in tokamaks

    International Nuclear Information System (INIS)

    A comprehensive analysis of the stability properties of the appropriate kinetically generalized form of MHD ballooning modes together with the usual trapped-particle drift modes is presented. The calculations are fully electromagnetic and include the complete dynamics associated with compressional ion acoustic waves. Trapped-particle effects along with all forms of collisionless dissipation are taken into account without approximations. The influence of collisions is estimated with a model Krook operator. Results from the application of this analysis to realistic tokamak operating conditions indicate that unstable short-wavelength modes with significant growth rates can extend from β = 0 to value above the upper ideal-MHD-critical-beta associated with the so-called second stability regime. Since the strength of the relevant modes appears to vary gradually with β, these results support a soft beta limit picture involving a continuous (rather than abrupt or hard) modification of anomalous transport already present in low-β-tokamaks. However, at higher beta the increasing dominance of the electromagnetic component of the perturbations indicated by these calculations could also imply significantly different transport scaling properties

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

    International Nuclear Information System (INIS)

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

  3. The first results of electrode biasing experiments in the IR-T1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Ghoranneviss, M; Salar Elahi, A; Mohammadi, S; Arvin, R, E-mail: salari_phy@yahoo.co [Plasma Physics Research Center, Science and Research Branch, Islamic Azad University, PO Box 14665-678, Tehran (Iran, Islamic Republic of)

    2010-09-15

    We report here the first results of our movable electrode biasing experiments performed in the IR-T1 tokamak. For this study, a movable electrode biasing system was designed, constructed and installed on the IR-T1 tokamak. A positive voltage was applied to an electrode inserted in the tokamak limiter. The plasma current, poloidal and radial components of the magnetic fields, loop voltage and diamagnetic flux in the absence and presence of the biased electrode were measured. Results of the improvement done to plasma equilibrium behaviour are compared and discussed in this paper.

  4. 3D simulation studies of tokamak plasmas using MHD and extended-MHD models

    International Nuclear Information System (INIS)

    The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω*i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  6. Design and realization of central control system for J-TEXT tokamak

    International Nuclear Information System (INIS)

    To keep the J-TEXT tokamak running in order, it needs a central control system to coordinate the operations of all the sub-systems. The J-TEXT tokamak central control system adopts industrial PC with some PCI control cards based on Windows and QNX operation systems. This control system has been finished the test and put into use in the first-round experiment of the J-TEXT tokamak successfully. The testing and operation show that this system is stable and reliable. (authors)

  7. Study of energy transport in Tore Supra Tokamak; Etude du transport de l`energie sur le Tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Guiziou, L.

    1995-12-18

    The goal of this thesis is to characterize the energy confinement and the heat transport in Tore Supra tokamak. The first chapter is an introduction to the different plasma confinement regimes: ohmic, low confinement and improved confinement regimes. The second chapter is devoted to the presentation of the different theoretical and empirical approaches about energy confinement and heat transport. In the third chapter an attempt of explanations for non-local transport phenomenons is given. A turbulence correlation length greater than the ionic Larmor radius seams to be a reasonable explanation. This theoretical study focusses on the possibility for modes coupling in a tokamak. This study tries to determine a radial correlation length considering the two principal coupling modes: toroidal and non-linear. Different transport regimes are discussed using an analytical model and considering the influence of one coupling with respect to the other. In chapter four, the measurements of current profiles and transport coefficients are presented. The codes used for the reconstruction of equilibrium and for the experimental determination of the diffusivity are briefly presented. In chapter five, experimental results of energy transport studies for Tore Supra plasmas are presented. The different modes are analysed in detail and the study focusses on the influence of magnetic shear in the improved confinement regime. Finally, the different parametric dependences of the electronic thermal diffusivity are compared to local transport models. 165 refs., 57 figs., 2 tabs., 2 appendix.

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

    International Nuclear Information System (INIS)

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

  9. Theoretical studies of lower hybrid current drive in the Alcator C tokamak

    International Nuclear Information System (INIS)

    Theoretical studies of lower hybrid current drive in the Alcator C tokamak using a simulation model is presented, the model incorporates a 1-D radial transport code to solve for the time evolution of the bulk plasma quantities

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

    International Nuclear Information System (INIS)

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

  11. Experimental device for the X-ray energetic distribution measurement in a tokamak plasma

    International Nuclear Information System (INIS)

    An experimental system to measure the X-ray spectrum in a tokamak plasma is described, emphasizing its characteristics: resolution, dead time and the pulse pile-up distortion effects on the X-ray spectra. (author)

  12. A novel flexible field-aligned coordinate system for tokamak edge plasma simulation

    CERN Document Server

    Leddy, Jarrod; Romanelli, Michele; Shanahan, Brendan; Walkden, Nick

    2016-01-01

    Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are "closed" (ie. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines begin to intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry...

  13. Electron heat transport in current carrying and currentless thermonuclear plasmas. Tokamaks and stellarators compared

    International Nuclear Information System (INIS)

    In the first experiment the plasma current in the RTP tokamak is varied. Here the underlying idea was to check whether at a low plasma current, transport in the tokamak resembles transport in stellarators more than at higher currents. Secondly, experiments have been done to study the relation of the diffusivity χ to the temperature and its gradient in both W7-AS and RTP. In this case the underlying idea was to find the explanation for the phenomenon observed in both tokamaks and stellarators that the quality of the confinement degrades when more heating is applied. A possible explanation is that the diffusivity increases with the temperature or its gradient. Whereas in standard tokamak and stellarator experiments the temperature and its gradient are strongly correlated, a special capability of the plasma heating system of W7-AS and RTP can force them to decouple. (orig.)

  14. Transformer recharge by lower hybrid current drive on HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    The results of transformer recharging experiments by means of lower hybrid current drive (LHCD) on HT-7 superconducting Tokamak are summarized and analyzed on the basis of a simple transformer circuit. A steady-state plasma can be maintained at about 50 kA by LHCD while the primary coils of Tokamak transformer is recharged. During the recharge phase, a constant plasma loop voltage about -0.05 V keeps for about 1.0 s and the magnetic flux of the Tokamak transformer changes from 0.645 to 0.615 Vs. A maximum recharging efficiency of 1.25% has been obtained at a line averaged electron density of 0.5 x 1019 m-3. In addition, full non-inductive current drive experiment has been realized on HT-7 Tokamak

  15. Radio-frequency wave trajectories for current drive in tokamak reactors

    International Nuclear Information System (INIS)

    Detailed ray tracing calculations were carried out for three modes of waveguide-launched radio-frequency waves for tokamak reactor parameters to evaluate their applicability for steady state current drive. The merits and demerits of each mode are discussed

  16. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    Science.gov (United States)

    Sager, G. T.; Wong, C. P. C.; Kapich, D. D.; McDonald, C. F.; Schleicher, R. W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.

  17. Radio-frequency wave trajectories for current drive in tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wong, K.L.; Ono, M.

    1982-12-01

    Detailed ray tracing calculations were carried out for three modes of waveguide-launched radio-frequency waves for tokamak reactor parameters to evaluate their applicability for steady state current drive. The merits and demerits of each mode are discussed.

  18. A procedure for the design of snowflake magnetic configurations in tokamaks

    International Nuclear Information System (INIS)

    This paper deals with the design of snowflake (SF) plasma configurations in tokamaks. The SF configuration represents a promising solution for the power exhaust and divertor design problem due to its ability to flare the scrape-off layer in the vicinity of the SF point. SF plasma configurations have been successfully achieved in tokamaks like Tokamak à Configuration Variable (TCV), DIII-D and National Spherical Torus Experiment (NSTX), and are under investigation for future tokamaks such as DEMO. The first attempts to determine such new plasma configurations have picked out the inherent difficulties in obtaining them with low coil currents and in controlling the associated equilibria against external disturbances and modeling errors. This paper presents a novel procedure based on the linearized model of the plasma for the design of an SF divertor configuration. Moreover, a procedure for the optimization of the poloidial field coil system is proposed. The effectiveness of the techniques is demonstrated with an application to DEMO. (paper)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

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