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Sample records for aditya tokamak research

  1. Diamagnetic flux measurement in Aditya tokamak

    SciTech Connect

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

    2010-12-15

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

  2. Diamagnetic flux measurement in Aditya tokamak.

    PubMed

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

    2010-12-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-06-01

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

  4. Forecasting disruptions in the ADITYA tokamak using neural networks

    NASA Astrophysics Data System (ADS)

    Sengupta, A.; Ranjan, P.

    2000-12-01

    A neural network technique has been used to predict disruptions in the ADITYA tokamak. A time series prediction method is employed whereby a series of past values of some time dependent quantity is used to predict its value in the future. The time varying observables used in the present work are the different diagnostic signals from four Mirnov probes, one soft X ray monitor and one Hα monitor. The predicted quantities are the same observables at some future time. The neural network is trained with the past values of the different diagnostic signals as inputs and the future values of the same quantities as targets. The trained neural network is used to forecast in a multistep sequence. This amounts to a prediction several time steps earlier. Very good prediction results have been obtained up to 8 ms earlier with little distortion of the signals and no appreciable time lag, a capability which is believed to be well suited to the task of on-line predictions of disruptions in ADITYA. As actual experimental signals are used, confidence regarding the performance of the neural network on hardware implementation is automatically ensured.

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

    SciTech Connect

    Sangwan, Deepak; Jha, Ratneshwar; Tanna, Rakesh L.

    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.

  6. Improvement of Plasma Performance with Lithium Wall Conditioning in Aditya Tokamak

    NASA Astrophysics Data System (ADS)

    B. Chowdhuri, M.; Manchanda, R.; Ghosh, J.; B. Bhatt, S.; Ajai, Kumar; K. Das, B.; A. Jadeja, K.; A. Raijada, P.; Manoj, Kumar; Banerjee, S.; Nilam, Ramaiya; Aniruddh, Mali; Ketan, M. Patel; Vinay, Kumar; Vasu, P.; Bhattacharyay, R.; L. Tanna, R.; Y. Shankara, Joisa; K. Atrey, P.; V. S. Rao, C.; Chenna Reddy, D.; K. Chattopadhyay, P.; Jha, R.; C. Saxena, Y.; Aditya Team

    2013-02-01

    Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ~100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ~40% to 50%) and a 20% to 30% decrease of Hα signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Zeff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.

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

    NASA Astrophysics Data System (ADS)

    Dhyani, Pravesh; Ghosh, Joydeep; Sathyanarayana, K.; Praveenlal, V. E.; Gautam, Pramila; Shah, Minsha; Tanna, R. L.; Kumar, Pintu; Chavda, C.; Patel, N. C.; Panchal, V.; Gupta, C. N.; Jadeja, K. A.; Bhatt, S. B.; Kumar, S.; Raju, D.; Atrey, P. K.; Joisa, S.; Chattopadhyay, P. K.; Saxena, Y. C.

    2014-10-01

    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.

  8. Modification of plasma flows with gas puff in the scrape-off layer of ADITYA tokamak

    SciTech Connect

    Sangwan, Deepak; Jha, Ratneshwar; Brotankova, Jana; Gopalkrishna, M. V.

    2013-06-15

    The parallel Mach numbers are measured at three locations in the scrape-off layer (SOL) plasma of ADITYA tokamak by using Mach probes. The flow pattern is constructed from these measurements and the modification of flow pattern is observed by introducing a small puff of working gas. In the normal discharge, there is an indication of shell structure in the SOL plasma flows, which is removed during the gas puff. The plasma parameters, particle flux and Reynolds stress are also measured in the normal discharge and in the discharge with gas puff. It is observed that Reynolds stress and Mach number are coupled in the near SOL region and decoupled in the far SOL region. The coupling in the near SOL region gets washed away during the gas puff.

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

    SciTech Connect

    Chowdhuri, M. B. 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.; Kumar, Ajay

    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.

  10. Transport driven plasma flows in the scrape-off layer of ADITYA Tokamak in different orientations of magnetic field

    SciTech Connect

    Sangwan, Deepak; Jha, Ratneshwar; Brotankova, Jana; Gopalkrishna, M. V.

    2014-06-15

    Parallel plasma flows in the scrape-off layer of ADITYA tokamak are measured in two orientations of total magnetic field. In each orientation, experiments are carried out by reversing the direction of the toroidal magnetic field and the plasma current. The transport-driven component is determined by averaging flow Mach numbers, measured in two directions of the toroidal magnetic field and the plasma current for the same orientation. It is observed that there is a significant transport-driven component in the measured flow and the component depends on the field orientation.

  11. Core-ion temperature measurement of the ADITYA tokamak using passive charge exchange neutral particle energy analyzer.

    PubMed

    Pandya, Santosh P; Ajay, Kumar; Mishra, Priyanka; Dhingra, Rajani D; Govindarajan, J

    2013-02-01

    Core-ion temperature measurements have been carried out by the energy analysis of passive charge exchange (CX) neutrals escaping out of the ADITYA tokamak plasma (minor radius, a = 25 cm and major radius, R = 75 cm) using a 45° parallel plate electrostatic energy analyzer. The neutral particle analyzer (NPA) uses a gas cell configuration for re-ionizing the CX-neutrals and channel electron multipliers (CEMs) as detectors. Energy calibration of the NPA has been carried out using ion-source and ΔE∕E of high-energy channel has been found to be ∼10%. Low signal to noise ratio (SNR) due to VUV reflections on the CEMs was identified during the operation of the NPA with ADITYA plasma discharges. This problem was rectified by upgrading the system by incorporating the additional components and arrangements to suppress VUV radiations and improve its VUV rejection capabilities. The noise rejection capability of the NPA was experimentally confirmed using a standard UV-source and also during the plasma discharges to get an adequate SNR (>30) at the energy channels. Core-ion temperature Ti(0) during flattop of the plasma current has been measured to be up to 150 eV during ohmically heated plasma discharges which is nearly 40% of the average core-electron temperature (typically Te(0) ∼ 400 eV). The present paper describes the principle of tokamak ion temperature measurement, NPA's design, development, and calibration along with the modifications carried out for minimizing the interference of plasma radiations in the CX-spectrum. Performance of the NPA during plasma discharges and experimental results on the measurement of ion-temperature have also been reported here.

  12. Study of Laser Produced Plasma of Limiter of the Aditya Tokomak for Detection of Molecular Bands

    NASA Astrophysics Data System (ADS)

    Rai, Awadhesh Kumar

    2016-06-01

    The tokamak wall protection is one of the prime concerns, and for this purpose, limiters are used. Graphite is commonly used as a limiter material and first wall material for complete coverage of the internal vacuum vessel surfaces of the tokamak. From the past few years, we are working to identify and quantify the impurities deposited on the different part of Aditya Tokamak in collaboration with the Scientists at Institute of Plasma Research, Ahmedabad, India using Laser Induced Breakdown Spectroscopy (LIBS) [1-3]. Laser induced breakdown spectroscopy (LIBS) spectra of limiter of Aditya Tokamak have been recorded in the spectral range of 200-900 nm in open atmosphere. Along with atomic and ionic spectral lines of the constituent elements of the limiter (1-3), LIBS spectra also give the molecular bands. When a high power laser beam is focused on the sample, laser induced plasma is produced on its surface. In early stage of the plasma Back ground continuum is dominated due to free-free or free-bound emission. Just after few nanoseconds the light from the plasma is dominated by ionic emission. Atomic emission spectra is dominated from the laser induced plasma during the first few microsecond after an ablation pulse where as molecular spectra is generated later when the plasma further cools down. For this purpose the LIBS spectra has been recorded with varying gate delay and gate width. The spectra of the limiter show the presence of molecular bands of CN and C2. To get better signal to background ratios of the molecular bands, different experimental parameters like gate delay, gate width, collection angle and collection point (spatial analysis off the plasama) of the plasma have been optimized. Thus the present paper deals with the variation of spectral intensity of the molecular bands with different experimental parameters. Keywords: Limiter, Molecular bands, C2, CN. References: 1. Proof-of-concept experiment for On-line LIBS Analysis of Impurity Layer Deposited on

  13. 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

    SciTech Connect

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

    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.

  14. 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.

    PubMed

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

    2015-12-01

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

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

    SciTech Connect

    Henning, C.D.; Logan, B.G.; Barr, W.L.; Bulmer, R.H.; Doggett, J.N.; Johnson, B.M.; Lee, J.D.; Hoard, R.W.; Miller, J.R.; Slack, D.S.

    1985-11-01

    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. (WRF)

  16. TIBER: tokamak ignition/burn experimental research

    SciTech Connect

    Henning, C.D.; Logan, B.G.; Barr, W.L.; Bulmer, R.H.; Doggett, J.N.; Johnston, B.M.; Hoard, R.W.; Lee, J.D.; Miller, J.R.; Slack, D.S.; Schultz, J.H.

    1985-11-01

    As part of a continuing effort by the Office of Fusion Energy to define an ignition experiment, a superconducting tokamak has been designed with thin neutron shielding and aggressive magnet and plasma parameters. By so minimizing the inner radial dimensions of the tokamak center post, coil, and shielding region, the plasma major radius is reduced, with a corresponding reduction in device costs. The peak nuclear-heating rate in the superconducting TF coils is 22 mW/cmT, which results in a steady heat load of 50 kW to the cryogenic system. Fast-wave, lower-hybrid heating would be used to induce a 10-MA current in a moderate density plasma. Then pellet fueling would raise the density to achieve ignition as the current decays in a few hundred seconds. Steady-state current drive in subignited conditions permits a 0.8 MW/mS average wall loading to study plasma and nuclear engineering effects. 10 refs., 6 figs., 3 tabs.

  17. Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak

    NASA Astrophysics Data System (ADS)

    Seo, Seong-Heon; Park, Jinhyung; Wi, H. M.; Lee, W. R.; Kim, H. S.; Lee, T. G.; Kim, Y. S.; Kang, Jin-Seob; Bog, M. G.; Yokota, Y.; Mase, A.

    2013-08-01

    Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6-54 GHz), V band (48-72 GHz), and W band (72-108 GHz) to measure the density up to 7 × 1019 m-3 when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.

  18. Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak.

    PubMed

    Seo, Seong-Heon; Park, Jinhyung; Wi, H M; Lee, W R; Kim, H S; Lee, T G; Kim, Y S; Kang, Jin-Seob; Bog, M G; Yokota, Y; Mase, A

    2013-08-01

    Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6-54 GHz), V band (48-72 GHz), and W band (72-108 GHz) to measure the density up to 7 × 10(19) m(-3) when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.

  19. Generation of multiple analog pulses with different duty cycles within VME control system for ICRH Aditya system

    NASA Astrophysics Data System (ADS)

    Joshi, Ramesh; Singh, Manoj; Jadav, H. M.; Misra, Kishor; Kulkarni, S. V.; ICRH-RF Group

    2010-02-01

    Ion Cyclotron Resonance Heating (ICRH) is a promising heating method for a fusion device due to its localized power deposition profile, a direct ion heating at high density, and established technology for high RF power generation and transmission at low cost. Multiple analog pulse with different duty cycle in master of digital pulse for Data acquisition and Control system for steady state RF ICRH System(RF ICRH DAC) to be used for operating of RF Generator in Aditya to produce pre ionization and second analog pulse will produce heating. The control system software is based upon single digital pulse operation for RF source. It is planned to integrate multiple analog pulses with different duty cycle in master of digital pulse for Data acquisition and Control system for RF ICRH System(RF ICRH DAC) to be used for operating of RF Generator in Aditya tokamak. The task of RF ICRH DAC is to control and acquisition of all ICRH system operation with all control loop and acquisition for post analysis of data with java based tool. For pre ionization startup as well as heating experiments using multiple RF Power of different powers and duration. The experiment based upon the idea of using single RF generator to energize antenna inside the tokamak to radiate power twise, out of which first analog pulse will produce pre ionization and second analog pulse will produce heating. The whole system is based on standard client server technology using tcp/ip protocol. DAC Software is based on linux operating system for highly reliable, secure and stable system operation in failsafe manner. Client system is based on tcl/tk like toolkit for user interface with c/c++ like environment which is reliable programming languages widely used on stand alone system operation with server as vxWorks real time operating system like environment. The paper is focused on the Data acquisition and monitoring system software on Aditya RF ICRH System with analog pulses in slave mode with digital pulse in

  20. Recent research work on the J-TEXT tokamak

    NASA Astrophysics Data System (ADS)

    Zhuang, G.; Gentle, K. W.; Rao, B.; Feng, X. D.; Chen, J.; Hu, Q. M.; Jin, W.; Hu, X. W.; Chen, Z. Y.; Wang, Z. J.; Ding, Y. H.; Zhang, M.; Chen, Z. P.; Yang, Z. J.; Gao, L.; Zhang, X. Q.; Cheng, Z. F.; Pan, Y.; Yu, K. X.; Huang, H.; the J-TEXT Team

    2013-10-01

    An overview of the recent research work on the J-TEXT tokamak over the last two years is presented. A series of experiments and simulations of the interaction between resonant magnetic perturbations (RMPs) and plasma were carried out on the J-TEXT tokamak. The results show that the m/n = 2/1 (m and n are the poloidal and toroidal mode numbers, respectively) mode locking is obtained with sufficiently large RMPs. And suppression of the m/n = 2/1 tearing mode by moderate magnetic perturbation amplitude is also observed. With experimental parameters as input, both mode locking and mode suppression by RMPs are simulated by nonlinear numerical modelling based on reduced magnetohydrodynamic equations. The simulations are in good agreement with the experimental observations. Density modulation using gas puffing is carried out on J-TEXT to evaluate the particle transport parameters in a typical J-TEXT discharge, including diffusion coefficient and convective velocity. Inverse sawtooth-like activity caused by neon gas injection is observed. The inverse sawtooth-like activity occurs only when the amount of neon impurity exceeds a threshold. Nevertheless, other impurities such as helium and argon cannot trigger such events. With the aid of a soft x-ray detector array, the runaway electron beam following disruptions is visible directly. A high-resolution far infrared polarimeter/interferometer, based on a three-wave technique, was developed and it observes the perturbations associated with sawtooth and tearing mode activities; the first result of the current density profile reconstruction is provided. An x-ray imaging crystal spectrometer is designed to receive the Kα line of Ar XVII and its satellites. The electron temperature obtained from line ratios of the W line to its satellites is 750 eV, and the ion temperature deduced from the Doppler broadening of the W line is 330 eV.

  1. [Fusion research/tokamak]. Final report, 1 May 1988--30 April 1994

    SciTech Connect

    1994-12-31

    The objectives of the Fusion Research Center Program are: (1) to advance /the transport studies of tokamaks, including the development and maintenance of the Magnetic Fusion Energy Database, and (2) to provide theoretical interpretation, modeling and equilibrium and stability studies for the text-upgrade tokamak. Work is described on five basic categories: (1) magnetic fusion energy database; (2) computational support and numerical modeling; (3) support for TEXT-upgrade and diagnostics; (4) transport studies; and (5) Alfven waves.

  2. Diamagnetic loop measurement in Korea Superconducting Tokamak Advanced Research machine.

    PubMed

    Bak, J G; Lee, S G; Kim, H S

    2011-06-01

    Diamagnetic loop (DL), which consists of two poloidal loops inside the vacuum vessel, is used to measure the diamagnetic flux during a plasma discharge in the Korea Superconducting Tokamak Advanced Research (KSTAR) machine. The vacuum fluxes in the DL signal can be compensated up to 0.1 mWb by using the coefficients, which are obtained from experimental investigations, in the vacuum flux measurements during vacuum shots under same operational conditions of magnetic coils for plasma experiment in the KSTAR machine. The maximum error in the diamagnetic flux measurement due to the errors of the coefficients was estimated as ∼0.22 mWb. From the diamagnetic flux measurements for the ohmically heated circular plasmas in the KSTAR machine, the stored energy agrees well with the estimated kinetic energy within the discrepancy of 25%. When the electron cyclotron heating, the neutral beam injection, and the ion cyclotron resonance heating are added to the ohmically heated limiter plasmas, the additional heating effects can be clearly observed from the increase of the stored energy evaluated in the DL measurement.

  3. Diamagnetic loop measurement in Korea Superconducting Tokamak Advanced Research machine

    SciTech Connect

    Bak, J. G.; Lee, S. G.; Kim, H. S.

    2011-06-15

    Diamagnetic loop (DL), which consists of two poloidal loops inside the vacuum vessel, is used to measure the diamagnetic flux during a plasma discharge in the Korea Superconducting Tokamak Advanced Research (KSTAR) machine. The vacuum fluxes in the DL signal can be compensated up to 0.1 mWb by using the coefficients, which are obtained from experimental investigations, in the vacuum flux measurements during vacuum shots under same operational conditions of magnetic coils for plasma experiment in the KSTAR machine. The maximum error in the diamagnetic flux measurement due to the errors of the coefficients was estimated as {approx}0.22 mWb. From the diamagnetic flux measurements for the ohmically heated circular plasmas in the KSTAR machine, the stored energy agrees well with the estimated kinetic energy within the discrepancy of 25%. When the electron cyclotron heating, the neutral beam injection, and the ion cyclotron resonance heating are added to the ohmically heated limiter plasmas, the additional heating effects can be clearly observed from the increase of the stored energy evaluated in the DL measurement.

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

    SciTech Connect

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

    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.

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

    PubMed

    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.

  6. An enhanced tokamak startup model

    NASA Astrophysics Data System (ADS)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

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

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

    NASA Astrophysics Data System (ADS)

    S. Codathe TCV Team

    2013-10-01

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

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

    SciTech Connect

    Wootton, A.J.

    1994-07-01

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

  9. Survey of nuclear fusion technology. A prospect analysis of Tokamak fusion research

    NASA Astrophysics Data System (ADS)

    Hwang, W. K.

    The fusion research evaluation model for analyzing various R and D sinarios, the trend analysis of Tokamak research, and the near-term technologies are discussed. A computer code, FUSREV, was developed. It consists of the plasma power model and the cost/benefit model. Since the state-of-the-arts which are expected to be achieve as the result of subproject R and D's can only be obtained in the form of probability distribution functions Monte Carlo method is employed. The test computation of the code shows acceptable results. However, FUSREV was continuously modified employing new models for both technology and economics.

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

    SciTech Connect

    Wootton, A.J.

    1995-08-01

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

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

    SciTech Connect

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

    2015-07-15

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

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

    PubMed

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

    2015-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

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

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

    SciTech Connect

    Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.; and others

    2014-11-15

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

  15. Analog integrator for the Korea superconducting tokamak advanced research magnetic diagnostics

    SciTech Connect

    Bak, J. G.; Lee, S. G.; Son, D.; Ga, E. M.

    2007-04-15

    An analog integrator, which automatically compensates an integrating drift, has been developed for the magnetic diagnostics in the Korea superconducting tokamak advanced research (KSTAR). The compensation of the drift is done by the analog to digital converter-register-digital to analog converter in the integrator. The integrator will be used in the equilibrium magnetic field measurements by using inductive magnetic sensors during a plasma discharge in the KSTAR machine. Two differential amplifiers are added to the signal path between each magnetic sensor and the integrator in order to improve the performance of the integrator because a long signal cable of 100 m will be used for the measurement in the KSTAR machine. In this work, the characteristics of the integrator with two differential amplifiers are experimentally investigated.

  16. Analog integrator for the Korea superconducting tokamak advanced research magnetic diagnostics

    NASA Astrophysics Data System (ADS)

    Bak, J. G.; Lee, S. G.; Son, D.; Ga, E. M.

    2007-04-01

    An analog integrator, which automatically compensates an integrating drift, has been developed for the magnetic diagnostics in the Korea superconducting tokamak advanced research (KSTAR). The compensation of the drift is done by the analog to digital converter-register-digital to analog converter in the integrator. The integrator will be used in the equilibrium magnetic field measurements by using inductive magnetic sensors during a plasma discharge in the KSTAR machine. Two differential amplifiers are added to the signal path between each magnetic sensor and the integrator in order to improve the performance of the integrator because a long signal cable of 100 m will be used for the measurement in the KSTAR machine. In this work, the characteristics of the integrator with two differential amplifiers are experimentally investigated.

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

    SciTech Connect

    Spampinato, P.T.

    1988-01-01

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

  18. Analog integrator for the Korea superconducting tokamak advanced research magnetic diagnostics.

    PubMed

    Bak, J G; Lee, S G; Son, D; Ga, E M

    2007-04-01

    An analog integrator, which automatically compensates an integrating drift, has been developed for the magnetic diagnostics in the Korea superconducting tokamak advanced research (KSTAR). The compensation of the drift is done by the analog to digital converter-register-digital to analog converter in the integrator. The integrator will be used in the equilibrium magnetic field measurements by using inductive magnetic sensors during a plasma discharge in the KSTAR machine. Two differential amplifiers are added to the signal path between each magnetic sensor and the integrator in order to improve the performance of the integrator because a long signal cable of 100 m will be used for the measurement in the KSTAR machine. In this work, the characteristics of the integrator with two differential amplifiers are experimentally investigated.

  19. Twenty Years of Research on the Alcator C-Mod Tokamak

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin

    2013-10-01

    Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

  20. Investigation of relativistic runaway electrons in electron cyclotron resonance heating discharges on Korea Superconducting Tokamak Advanced Research

    SciTech Connect

    Kang, C. S.; Lee, S. G.

    2014-07-15

    The behavior of relativistic runaway electrons during Electron Cyclotron Resonance Heating (ECRH) discharges is investigated in the Korea Superconducting Tokamak Advanced Research device. The effect of the ECRH on the runaway electron population is discussed. Observations on the generation of superthermal electrons during ECRH will be reported, which will be shown to be consistent with existing theory for the development of a superthermal electron avalanche during ECRH [A. Lazaros, Phys. Plasmas 8, 1263 (2001)].

  1. Bibliography of fusion product physics in tokamaks

    SciTech Connect

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

    1989-09-01

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

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

    SciTech Connect

    none,

    1996-05-31

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

  3. Role of neutral gas in scrape-off layer tokamak plasma

    SciTech Connect

    Bisai, N.; Jha, R.; Kaw, P. K.

    2015-02-15

    Neutral gas in scrape-off layer of tokamak plasma plays an important role as it can modify the plasma turbulence. In order to investigate this, we have derived a simple two-dimensional (2D) model that consists of electron continuity, quasi-neutrality, and neutral gas continuity equations using neutral gas ionization and charge exchange processes. Simple 1D profile analysis predicts neutral penetration depth into the plasma. Growth rate obtained from the linear theory has been presented. The 2D model equations have been solved numerically. It is found that the neutral gas reduces plasma fluctuations and shifts spectrum of the turbulence towards lower frequency side. The neutral gas fluctuation levels have been presented. The numerical results have been compared with Aditya tokamak experiments.

  4. Design of a single-channel millimeter-wave interferometer system for Korea Superconducting Tokamak Advanced Research

    NASA Astrophysics Data System (ADS)

    Nam, Y. U.; Cheon, M. S.; Kwon, M.; Hwang, Y. S.

    2003-03-01

    A simple single-channel horizontal millimeter-wave interferometer has been designed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR). To measure line integrated plasma densities of 2×1019 m-2 in the initial phase of the KSTAR, Gunn oscillator frequency of 280 GHz has been chosen to optimize errors due to both vibration on the beam path and refraction in the plasma. To reduce the free propagation length of the probing beam and to obtain small beam width on the vacuum windows, a retractable cassette system for deep positioning of the diagnostic system has been designed, where microwave parts are located as close as possible to the tokamak with a shielding box. A beam focusing system with concave reflecting mirrors has been designed on the cassette and on the inner wall of the tokamak to reduce beam losses and to minimize beam width in the plasma. The estimated total transmission loss is about 25 dB, and beam widths are reduced significantly in the range of 20-50 mm.

  5. Recent progress of magnetic reconnection research in the MAST spherical tokamak

    NASA Astrophysics Data System (ADS)

    Tanabe, Hiroshi

    2016-10-01

    In the last three years, magnetic reconnection research in the MAST spherical tokamak achieved major progress by use of new 32 chord ion Doppler tomography, 130 channel YAG- and 300 channel Ruby-Thomson scattering diagnostics. In addition to the significant plasma heating up to 1 keV, detailed full temperature profile measurements including the diffusion region have been achieved for the first time. 2D imaging measurements of Ti and Te profiles have revealed that magnetic reconnection mostly heats ions globally in the downstream region of outflow jet and electrons locally at the X-point. The higher toroidal field in MAST (Bt > 0.3 T) strongly inhibits cross-field thermal transport scaling as 1 /Bt2 and the characteristic peaked Te profile at the X point is sustained on a millisecond time scale. In contrast, ions are mostly heated in the downstream region of outflow acceleration inside the current sheet width (c /ωpi 0.1 m) and around the stagnation point formed by reconnected flux mostly by viscosity dissipation and shock-like compressional damping of the outflow jet. Toroidal confinement also contributes to the characteristic Ti profile, forming a ring structure aligned with the closed flux surface. There is an effective confinement of the downstream thermal energy due to a thick layer of reconnected flux. The characteristic structure is sustained for longer than an ion-electron energy relaxation time (τeiE 4 - 11 ms) and the energy exchange between ions and electrons contributes to the bulk electron heating in the downstream region. The toroidal guide field mostly contributes to the formation of a localized electron heating structure at the X-point but not to bulk ion heating downstream. This work was supported by Grant-in-Aid for Scientific Research 15H05750, 15K14279 and 15K20921.

  6. DIII-D research towards resolving key issues for ITER and steady-state tokamaks

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; the DIII-D Team

    2013-10-01

    The DIII-D research program is addressing key ITER research needs and developing the physics basis for future steady-state tokamaks. Pellet pacing edge-localized mode (ELM) control in the ITER configuration reduces ELM energy loss in proportion to 1/fpellet by inducing ELMs at up to 12× the natural ELM rate. Complete suppression of ELMs with resonant magnetic perturbations has been extended to the q95 expected for ITER baseline scenario discharges, and long-duration ELM-free QH-mode discharges have been produced with ITER-relevant co-current neutral-beam injection (NBI) using external n = 3 coils to generate sufficient counter-Ip torque. ITER baseline discharges at βN ˜ 2 and scaled NBI torque have been maintained in stationary conditions for more than four resistive times using electron cyclotron current drive (ECCD) for tearing mode suppression and disruption avoidance; active tracking with steerable launchers and feedback control catch these modes at small amplitude, reducing the ECCD power required to suppress them. Massive high-Z gas injection into disruption-induced 300-600 kA 20 MeV runaway electron (RE) beams yield dissipation rates ˜10× faster than expected from e-e collisions and demonstrate the possibility of benign dissipation of such REs should they occur in ITER. Other ITER-related experiments show measured intrinsic plasma torque in good agreement with a physics-based model over a wide range of conditions, while first-time main-ion rotation measurements show it to be lower than expected from neoclassical theory. Core turbulence measurements show increased temperature fluctuations correlated with sharply enhanced electron transport when \

  7. Overview of the recent research on the J-TEXT tokamak

    NASA Astrophysics Data System (ADS)

    Zhuang, G.; Gentle, K. W.; Diamond, Patrick; Chen, J.; Rao, B.; Wang, L.; Zhao, Kaijun; Han, Sanghee; Shi, Yuejiang; Ding, Y. H.; Chen, Z. Y.; Hu, X. W.; Wang, Z. J.; Yang, Z. J.; Chen, Z. P.; Cheng, Z. F.; Gao, L.; Zhang, X. Q.; Zhang, M.; Yu, K. X.; Pan, Y.; Huang, H.; the J-TEXT Team

    2015-10-01

    The experimental research over last two years on the J-TEXT tokamak is summarized and presented in the paper. The high-performance polarimeter-interferometer developed on J-TEXT, aiming to measure electron density and Faraday angle simultaneously, has time response up to 1 µs, phase resolution <0.1° and spatial resolution ∼3 cm. Such high resolution permits investigations of fast equilibrium dynamics as well as magnetic and density perturbations associated with magnetohydrodynamic instabilities. Particle transport due to the sawtooth crashes is analysed. The sawteeth only partially flatten the core density profile and recovery between crashes implies an inward pinch velocity extending to the centre. The resonant magnetic perturbation (RMP) system on J-TEXT can generate a rotating helical field perturbation with a maximum rotation frequency up to 6 kHz, and dominant resonant modes of m/n = 2/1, 3/1 or 1/1. It is found that tearing modes can be easily locked and then rotate together with a rotating RMP. The effects of RMPs on plasma flows and fluctuations are studied with Langmuir probe arrays at the plasma edge. The toroidal velocity increases and the radial electric field decreases with RMP coil current when the RMP current is no more than 5 kA. When the RMP current reaches 6 kA, the toroidal velocity profile becomes flattened near the last closed flux surface. The geodesic acoustic mode is damped in most of the edge region, while the low frequency zonal flow is damped inside the islands, but increases at its boundary.

  8. Thermal instability theory analysis of multifaceted asymmetric radiation from the edge (MARFE) in Tokamak Experiment for Technology Oriented Research (TEXTOR)

    SciTech Connect

    Kelly, F. A.; Stacey, W. M.; Rapp, J.; Brix, M.

    2001-07-01

    The density limits for a series of shots in TEXTOR [Tokamak Experiment for Technology Oriented Research, E. Hintz, P. Bogen, H. A. Claa{ss}en , in Contributions to High-Temperature Plasma Physics, edited by K. H. Spatschek and J. Uhlenbusch (Akademie Verlag, Berlin, 1994, p. 373)], over a range of heating powers, that ended in multifaceted asymmetric radiation from the edge (MARFE) have been analyzed within the context of thermal instability theory. The prediction of MARFE onset agrees with observation to within the experimental uncertainty.

  9. Summary discussion: An integrated advanced tokamak reactor

    SciTech Connect

    Sauthoff, N.R.

    1994-12-31

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ``figures of merit`` for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept.

  10. Texas Experimental Tokamak

    SciTech Connect

    Wootton, A.J.

    1993-04-01

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

  11. ECH tokamak

    SciTech Connect

    Firestone, M.A.; Mau, T.K.; Conn, R.W.

    1985-04-01

    A small steady-state tokamak capable of producing power in the 100 to 300 MWe range and relying on electron cyclotron RF heating (ECH) for both heating and current drive is described. Working in the first MHD stability regime for tokamaks, the approach adheres to the recently discovered maximum beta limit. An appropriate figure of merit is the ratio of the fusion power to absorbed RF power. Efficient devices are feasible at both small and large values of fusion power, thereby pointing to a development path for an attractive commercial fusion reactor.

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

    PubMed

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

    2016-04-01

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

  13. Research on the effect of resonant magnetic perturbations on disruption limit in J-TEXT tokamak

    NASA Astrophysics Data System (ADS)

    Hu, Qiming; Wang, Nengchao; Yu, Q.; Ding, Yonghua; Rao, Bo; Chen, Zhipeng; Jin, Hai

    2016-02-01

    The effect of resonant magnetic perturbations (RMPs) on high density limit and low-q limit discharges is studied in J-TEXT tokamak. It is found that a moderate amplitude of applied RMPs increases the density limit and delays the disruption, and the disruption precursor is suppressed with a slight reduction in toroidal plasma rotation. A too large amplitude of RMPs, however, leads to a lower density limit and an earlier disruption. For the optimal RMPs amplitude, the disruption is delayed by about 150 ms. For low-q discharges, the applied static RMP can either lower the limit of the edge safety factor q a from 2.15 to nearly 2.0 or sustain stable plasma discharge with q a  =  2.1, and the disruptive precursor is also suppressed.

  14. The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    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.

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

    NASA Astrophysics Data System (ADS)

    Huang, Zhi-Hui; Yan, Long-Wen; Tomita, Yukihiro; Feng, Zhen; Cheng, Jun; Hong, Wen-Yu; Pan, Yu-Dong; Yang, Qing-Wei; Duan, Xu-Ru

    2015-02-01

    The investigation of dust particle characteristics in fusion devices has become more and more imperative. In the HL-2A tokamak, the morphologies and compositions of dust particles are analyzed by using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDX) with mapping. The results indicate that the sizes of dust particles are in a range from 1 μm to 1 mm. Surprisingly, stainless steel spheres with a diameter of 2.5 μm-30 μm are obtained. The production mechanisms of dust particles include flaking, disintegration, agglomeration, and arcing. In addition, dynamic characteristics of the flaking dust particles are observed by a CMOS fast framing camera and simulated by a computer program. Both of the results display that the ion friction force is dominant in the toroidal direction, while the centrifugal force is crucial in the radial direction. Therefore, the visible dust particles are accelerated toriodally by the ion friction force and migrated radially by the centrifugal force. The averaged velocity of the grain is on the order of ˜ 100 m/s. These results provide an additional supplement for one of critical plasma-wall interaction (PWI) issues in the framework of the International Thermonuclear Experimental Reactor (ITER) programme. Project supported by the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2014GB107000 and 2013GB112008), the National Natural Science Foundation of China (Grant Nos. 11320101005, 11175060, 11375054, and 11075046), and the China-Korean Joint Foundation (Grant No. 2012DFG02230).

  16. Numerical tokamak turbulence project (OFES grand challenge)

    SciTech Connect

    Beer, M; Cohen, B I; Crotinger, J; Dawson, J; Decyk, V; Dimits, A M; Dorland, W D; Hammett, G W; Kerbel, G D; Leboeuf, J N; Lee, W W; Lin, Z; Nevins, W M; Reynders, J; Shumaker, D E; Smith, S; Sydora, R; Waltz, R E; Williams, T

    1999-08-27

    The primary research objective of the Numerical Tokamak Turbulence Project (NTTP) is to develop a predictive ability in modeling turbulent transport due to drift-type instabilities in the core of tokamak fusion experiments, through the use of three-dimensional kinetic and fluid simulations and the derivation of reduced models.

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

    SciTech Connect

    Lee, H. Y.; Hong, J. H.; Jang, J. H.; Park, J. S.; Choe, Wonho; Hahn, S. H.; Bak, J. G.; Lee, J. H.; Ko, W. H.; Lee, K. D.; Lee, S. H.; Lee, H. H.; Juhn, J.-W.; Kim, H. S.; Yoon, S. W.; Han, H.; Ghim, Y.-C.

    2015-12-15

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

  18. PPPL tokamak program

    SciTech Connect

    Furth, H.P.

    1984-10-01

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

  19. Tokamak Systems Code

    SciTech Connect

    Reid, R.L.; Barrett, R.J.; Brown, T.G.; Gorker, G.E.; Hooper, R.J.; Kalsi, S.S.; Metzler, D.H.; Peng, Y.K.M.; Roth, K.E.; Spampinato, P.T.

    1985-03-01

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

  20. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  1. High-beta steady-state research and future directions on the Japan Atomic Energy Research Institute Tokamak-60 Upgrade and the Japan Atomic Energy Research Institute Fusion Torus-2 Modified

    NASA Astrophysics Data System (ADS)

    Ishida, S.; JT-60 Team, JFT-2M Group

    2004-05-01

    In the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U), a high- β p ELMy H-mode (high-poloidal-beta high-confinement-mode with edge localized mode) plasma was sustained with β N ˜2.7 for 7.4 s. Real-time neoclassical tearing mode (NTM) stabilization system was established and effective NTM suppression by early electron cyclotron (EC) wave injection was demonstrated. High fusion triple product of n i (0)τ E T i (0)=3.1×10 20 keVṡsṡm -3 was achieved using the negative-ion based neutral beam current drive with β N ˜2.5 and the bootstrap current fraction f BS ˜50%. In a hot electron regime, a high electron cyclotron current drive efficiency of 4.2×10 18 A/W/m 2 was achieved at T e ˜21 keV . Innovative current start-up scenario produced a current hole plasma with a very high f BS ˜90%. No accumulation of helium and carbon impurities was observed for internal transport barrier (ITB) plasmas. While argon impurity was accumulated, EC injection effectively exhausted it across ITB. In a regime of ELM disappearance, a clear correlation between the ELM frequency and the toroidal velocity at pedestal was observed. In the Japan Atomic Energy Research Institute Fusion Torus-2 Modified (JFT-2M), high beta plasmas were produced with full ferritic inside wall up to β N =3.3, where high recycling steady H-mode discharges were developed up to β N H 89 P ˜6 at n e /n GW ˜0.7-1.0 with ITB. JT-60U started long pulse experiment in late 2003 and JFT-2M will conduct wall stabilization experiment in early 2004. The modification of JT-60 to a fully superconducting coil tokamak is regarded as the national centralized tokamak facility program to accomplish the high beta steady-state research in a collisionless regime.

  2. Do spherical tokamaks have a thermonuclear future?

    NASA Astrophysics Data System (ADS)

    Mirnov, S. V.

    2012-12-01

    This work has been initiated by the publication of a review by B.V.Kuteev et al., "Intense Fusion Neutron Sources" [Plasma Physics Reports 36, 281 (2010)]. It is stated that the key thesis of the above review that a spherical tokamak can be recommended for research neutron sources and for demonstration hybrid systems as an alternative to expensive "classical" tokamaks of the JET and ITER type is inconsistent. The analysis of the experimental material obtained during the last 10 years in the course of studies on the existing spherical tokamaks shows that the TIN-ST fusion neutron source spherical tokamak proposed by the authors of the review and intended, according to the authors' opinion, to replace "monsters" in view of its table-top dimensions (2 m3) and laboratory-level energetics cannot be transformed into any noticeable stationary megawatt-power neutron source competing with the existing classical tokamaks (in particular, with JET with its quasi-steady DT fusion power at a level of 5 MW). Namely, the maximum plasma current in the proposed tokamak will be not 3 MA, as the authors suppose erroneously, but, according to the present-day practice of spherical tokamaks, within 0.6-0.7 MA, which will lead to a reduction on the neutron flux by two to three orders of magnitude from the expected 5 MW. The possibility of the maintenance of the stationary process itself even in such a "weakened" spherical tokamak is very doubtful. The experience of the largest existing devices of this type (such as NSTX and MAST) has shown that they are incapable of operating even in a quasi-steady operating mode, because the discharge in them is spontaneously interrupted about 1 s after the beginning of the current pulse, although its expected duration is of up to 5 s. The nature of this phenomenon is the subject of further study of the physics of spherical tokamaks. This work deals with a critical analysis of the available experimental data concerning such tokamaks and a discussion of

  3. Tokamak ARC damage

    SciTech Connect

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

    1985-01-01

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

  4. Tokamak reactor studies

    SciTech Connect

    Baker, C.C.

    1981-01-01

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

  5. Options for commercial tokamaks

    SciTech Connect

    Dabiri, A.E.; Keeton, D.C.; Thomson, S.L.

    1986-07-01

    Systems studies have been performed at the Fusion Engineering Design Center (FEDC) to assess commercial tokamak options. One study investigates the economics of high-beta operation and determines an optimum operating range of 10 to 20% beta, with a corresponding neutron wall loading of 6 to 8 MW/m/sup 2/. A second study determines conditions under which small, low-power tokamaks can be economically combined into a 1200-MW(electric) multiplex power plant. The results of these studies have directed future efforts at the FEDC toward a high-beta, tokamak design using a modular maintenance configuration.

  6. Overview of the National Centralized Tokamak programme

    NASA Astrophysics Data System (ADS)

    Kikuchi, M.; Tamai, H.; Matsukawa, M.; Fujita, T.; Takase, Y.; Sakurai, S.; Kizu, K.; Tsuchiya, K.; Kurita, G.; Morioka, A.; Hayashi, N.; Miura, Y.; Itoh, S.; Bialek, J.; Navratil, G.; Ikeda, Y.; Fujii, T.; Kurihara, K.; Kubo, H.; Kamada, Y.; Miya, N.; Suzuki, T.; Hamamatsu, K.; Kawashima, H.; Kudo, Y.; Masaki, K.; Takahashi, H.; Takechi, M.; Akiba, M.; Okuno, K.; Ishida, S.; Ichimura, M.; Imai, T.; Hashizume; Miura, Y. M.; Horiike, H.; Kimura, A.; Tsutsui, H.; Matsuoka, M.; Uesugi, Y.; Sagara, A.; Nishimura, A.; Shimizu, A.; Sakamoto, M.; Nakamura, K.; Sato, K.; Okano, K.; Ida, K.; Shimada, H. R.; Kishimoto, Y.; Azechi, H.; Tanaka, S.; Yatsu, K.; Yoshida, N.; Inutake, M.; Fujiwara, M.; Inoue, N.; Hosogane, N.; Kuriyama, M.; Ninomiya, H.

    2006-03-01

    An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.

  7. Modular tokamak magnetic system

    DOEpatents

    Yang, Tien-Fang

    1988-01-01

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

  8. Bifurcated helical core equilibrium states in tokamaks

    NASA Astrophysics Data System (ADS)

    Cooper, W. A.; Chapman, I. T.; Schmitz, O.; Turnbull, A. D.; Tobias, B. J.; Lazarus, E. A.; Turco, F.; Lanctot, M. J.; Evans, T. E.; Graves, J. P.; Brunetti, D.; Pfefferlé, D.; Reimerdes, H.; Sauter, O.; Halpern, F. D.; Tran, T. M.; Coda, S.; Duval, B. P.; Labit, B.; Pochelon, A.; Turnyanskiy, M. R.; Lao, L.; Luce, T. C.; Buttery, R.; Ferron, J. R.; Hollmann, E. M.; Petty, C. C.; van Zeeland, M.; Fenstermacher, M. E.; Hanson, J. M.; Lütjens, H.

    2013-07-01

    Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak à configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma-vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.

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

    SciTech Connect

    Strait, E. J.; Park, J. -K.; Marmar, E. S.; Ahn, J. -W.; Berkery, J. W.; Burrell, K. H.; Canik, J. M.; Delgado-Aparicio, L.; Ferraro, N. M.; Garofalo, A. M.; Gates, D. A.; Greenwald, M.; Kim, K.; King, J. D.; Lanctot, M. J.; Lazerson, S. A.; Liu, Y. Q.; Lore, J. D.; Menard, J. E.; Nazikian, R.; Shafer, M. W.; Paz-Soldan, C.; Reiman, A. H.; Rice, J. E.; Sabbagh, S. A.; Sugiyama, L.; Turnbull, A. D.; Volpe, F.; Wang, Z. R.; Wolfe, S. M.

    2014-09-30

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

  10. Electron cyclotron emission diagnostics on KSTAR tokamak

    SciTech Connect

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

    2010-10-15

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

  11. Advanced commercial tokamak study

    SciTech Connect

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs.

  12. Parametric dependence of density limits in the Tokamak Experiment for Technology Oriented Research (TEXTOR): Comparison of thermal instability theory with experiment

    NASA Astrophysics Data System (ADS)

    Kelly, F. A.; Stacey, W. M.; Rapp, J.

    2001-11-01

    The observed dependence of the TEXTOR [Tokamak Experiment for Technology Oriented Research: E. Hintz, P. Bogen, H. A. Claassen et al., Contributions to High Temperature Plasma Physics, edited by K. H. Spatschek and J. Uhlenbusch (Akademie Verlag, Berlin, 1994), p. 373] density limit on global parameters (I, B, P, etc.) and wall conditioning is compared with the predicted density limit parametric scaling of thermal instability theory. It is necessary first to relate the edge parameters of the thermal instability theory to n¯ and the other global parameters. The observed parametric dependence of the density limit in TEXTOR is generally consistent with the predicted density limit scaling of thermal instability theory. The observed wall conditioning dependence of the density limit can be reconciled with the theory in terms of the radiative emissivity temperature dependence of different impurities in the plasma edge. The thermal instability theory also provides an explanation of why symmetric detachment precedes radiative collapse for most low power shots, while a multifaceted asymmetric radiation from the edge MARFE precedes detachment for most high power shots.

  13. Completely bootstrapped tokamak

    SciTech Connect

    Weening, R.H. ); Boozer, A.H. )

    1992-01-01

    Numerical simulations of the evolution of large-scale magnetic fields have been developed using a mean-field Ohm's law. The Ohm's law is coupled to a {Delta}{prime} stabilty analysis and a magnetic island growth equation in order to simulate the behavior of tokamak plasmas that are subject to tearing modes. In one set of calculations, the magnetohydrodynamic (MHD)-stable regime of the tokamak is examined via the construction of an {ital l}{sub {ital i}} -{ital q}{sub {ital a}} diagram. The results confirm previous calculations that show that tearing modes introduce a stability boundary into the {ital l}{sub {ital i}} -{ital q}{sub {ital a}} space. In another series of simulations, the interaction between tearing modes and the bootstrap current is investigated. The results indicate that a completely bootstrapped tokamak may be possible, even in the absence of any externally applied loop voltage or current drive.

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

    NASA Astrophysics Data System (ADS)

    Azizov, E. A.

    2012-02-01

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

  15. Steady-state sustainment of high-{beta} plasmas through stability control in Japan Atomic Energy Research Institute Tokamak-60 Upgrade

    SciTech Connect

    Isayama, A.

    2005-05-15

    Recent results from steady-state sustainment of high-{beta} plasma experiments in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) tokamak [A. Kitsunezaki et al., Fusion Sci. Technol. 42, 179 (2002)] are described. Extension of discharge duration to 65 s (formerly 15 s) has enabled physics research with long time scale. In long-duration high-{beta} research, the normalized beta {beta}{sub N}=2.5, which is comparable to that in the steady-state operation in International Thermonuclear Experimental Reactor (ITER) [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Controlled Fusion 44, 519 (2002)], has been sustained for about 15 s with confinement enhancement factor H{sub 89PL} above 2, where the duration is about 80 times energy confinement time and {approx}10 times current diffusion time ({tau}{sub R}). In the scenario aiming at longer duration with {beta}{sub N}{approx}1.9, which is comparable to that in the ITER standard operation scenario, duration has been extended to 24 s ({approx}15{tau}{sub R}). Also, from the viewpoint of collisionality and Larmor radius of the plasmas, these results are obtained in the ITER-relevant regime with a few times larger than the ITER values. No serious effect of current diffusion on instabilities is observed in the region of {beta}{sub N} < or approx. 2.5, and in fact neoclassical tearing modes (NTMs), which limit the achievable {beta} in the stationary high-{beta}{sub p} H-mode discharges, are suppressed throughout the discharge. In high-{beta} research with the duration of several times {tau}{sub R}, a high-{beta} plasma with {beta}{sub N}{approx}2.9-3 has been sustained for 5-6 s with two scenarios for NTM suppression: (a) NTM avoidance by modification of pressure and current profiles, and (b) NTM stabilization with electron cyclotron current drive (ECCD)/electron cyclotron heating (ECH). NTM stabilization with the second harmonic X-mode ECCD/ECH has been performed, and it is found that EC current

  16. UCLA Tokamak Program Close Out Report.

    SciTech Connect

    Taylor, Robert John

    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.

  17. Modular tokamak configuration

    SciTech Connect

    Thomson, S.L.

    1985-01-01

    This report is concerned with the modular tokamak configuration, and presents information on the following topics: modularity; external vacuum boundary; vertical maintenance; combined reactor building/biological shield with totally remote maintenance; independent TF coils; minimum TF coil bore; saddle PF coils; and heat transport system in bore.

  18. Spontaneous generation of rotation in tokamak plasmas

    SciTech Connect

    Parra Diaz, Felix

    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.

  19. Development of a free-boundary tokamak equilibrium solver for advanced study of tokamak equilibria

    NASA Astrophysics Data System (ADS)

    Jeon, Young Mu

    2015-09-01

    A free-boundary Tokamak equilibrium solver (TES), developed for advanced study of tokamak equilibra, is described with two distinctive features. One is a generalized method to resolve the intrinsic axisymmetric instability, which is encountered in all equilibrium calculations with a freeboundary condition. The other is an extension to deal with a new divertor geometry such as snowflake or X divertors. For validations, the uniqueness of a solution is confirmed by the independence of variations in the computational domain, the mathematical correctness and accuracy of equilibrium profiles are checked by using a direct comparison with an analytic equilibrium known as a generalized Solov'ev equilibrium, and the governing force balance relation is tested by examining the intrinsic axisymmetric instabilities. As an application of an advanced equilibrium study, a snow-flake divertor configuration that requires a second-order zero of the poloidal magnetic flux is discussed in the circumstance of the Korea superconducting tokamak advanced research (KSTAR) coil system.

  20. Improvement of tokamak performance by injection of electrons

    SciTech Connect

    Ono, Masayuki

    1992-12-01

    Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas.

  1. Energy confinement in tokamaks

    SciTech Connect

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston.

  2. Magnetic confinement experiment -- 1: Tokamaks

    SciTech Connect

    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.

  3. Resistive edge mode instability in stellarator and tokamak geometries

    NASA Astrophysics Data System (ADS)

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-01

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  4. Resistive edge mode instability in stellarator and tokamak geometries

    SciTech Connect

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-15

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

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

    SciTech Connect

    1995-06-12

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

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

    SciTech Connect

    1992-08-01

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

  7. Edge turbulence in tokamaks

    NASA Astrophysics Data System (ADS)

    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.

  8. Dust Measurements in Tokamaks

    SciTech Connect

    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.

  9. ADX - Advanced Divertor and RF Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    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.

  10. Tokamak building-design considerations for a large tokamak device

    SciTech Connect

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release.

  11. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks

    SciTech Connect

    Scharer, J.E.

    1992-01-01

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  12. Bootstrap current in a tokamak

    SciTech Connect

    Kessel, C.E.

    1994-03-01

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

  13. Filamentation in tokamaks

    SciTech Connect

    Cardozo, N.J.; Barth, C.J.; Chu, C.C.; Lok, J.; Montvai, A.; Oomens, A.A.; Peters, M.; Pijper, F.J.; de Rover, M.; Schueller, F.C.; Steenbakkers, M.F.; RTP team

    1995-09-01

    The relevance of a nest of toroidal flux surfaces as a paradigm of the magnetic topology of a tokamak plasma is challenged. High resolution Thomson scattering measurements of electron temperature and density in RTP show several hot filaments in the plasma center and sharp gradients near the sawtooth inversion radius and structures outside the sawtooth region under central ECH. In ohmic plasmas, too, the pressure and temperature profiles show significant bumps. These measurements give evidence of a complex magnetic topology. Transport in a medium with spatially strongly varying diffusivity is considered. It is shown that macroscopic transport is determined by the microscopic structure: a transport theory must predict this structure and the diffusivity in the insulating regions, while the {open_quote}turbulent{close_quote} diffusivity is irrelevant. A numerical approach to equilibria with broken surfaces is presented. {copyright} {ital 1995 American Institute of Physics.}

  14. The Thor tokamak experiment

    NASA Astrophysics Data System (ADS)

    Argenti, L.; Bonizzoni, G.; Cirant, S.; Corti, S.; Grosso, G.; Lampis, G.; Rossi, L.; Carretta, U.; Jacchia, A.; de Luca, F.

    1981-06-01

    The principle characteristics of plasma discharges produced in Thor tokamak experiments are discussed. The equilibrium and stability characteristics of the plasma produced are considered, with attention given to the density limits and critical streaming parameter for stable operation. The temporal evolution of the main plasma parameters, including electron density, electron temperature distribution, hard X-ray emission from suprathermal electrons, neutral gas influx, plasma density and Ohmic heating efficiency, is then examined, with particular emphasis on means used to control the electron runaway. The results achieved are noted to have demonstrated the possibility of controlling both plasma equilibrium and discharge regime, and further improvements expected by the use of more efficient preionization, gas puffing and feedback poloidal control of column position are indicated.

  15. DIII-D research operations

    SciTech Connect

    Baker, D.

    1993-05-01

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

  16. Ripple-induced energetic particle loss in tokamaks

    NASA Astrophysics Data System (ADS)

    White, R. B.; Goldston, R. J.; Redi, M. H.; Budny, R. V.

    1996-08-01

    The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetic orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor [R. Hawryluk, Plasma Phys. Controlled Fusion 33, 1509 (1991)] and International Thermonuclear Experimental Reactor [K. Tomabechi, Proceedings of the 12th International Conference on Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1989), Vol. 3, p. 214] equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks.

  17. Texas Experimental Tokamak. Technical progress report, April 1990--April 1993

    SciTech Connect

    Wootton, A.J.

    1993-04-01

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

  18. Tokamak physics experiment: Diagnostic windows study

    SciTech Connect

    Merrigan, M.; Wurden, G.A.

    1995-11-01

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

  19. Overview of the Compact Ignition tokamak

    SciTech Connect

    Flanagan, C. A.; Peng, Yueng Kay Martin

    1986-01-01

    The Compact Ignition Tokamak (CIT) mission is to achieve ignition and provide the capability to experimentally study burning plasma behavior. A national team has developed a baseline concept including definition of the necessary research and development. The baseline concept satisfies the physics performance objectives established for the project and complies with defined design specifications. To ensure that the mission is achieved, the design requires large magnetic fields on axis (10 T) and use of large plasma currents (10 MA). The design is capable of accommodating significant auxiliary heating to enter the ignited regime. The CIT is designed to operate in plasma parameter regimes that a are directly relevant to future fusion power reactors.

  20. Moving Divertor Plates in a Tokamak

    SciTech Connect

    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.

  1. Resistive instabilities in tokamaks

    SciTech Connect

    Rutherford, P.H.

    1985-10-01

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

  2. Status of tokamak experiments

    SciTech Connect

    Wolf, G.H.

    1996-03-01

    Plasma-wall interaction, heat removal and ash exhaust have emerged as the dominant problems still to be solved in order to achieve ignition and - even more difficult - to maintain a state of self-sustained thermo-nuclear burn. This is of particular delicacy, since those operational regimes which yield the best energy confinement correspond to an even better particle confinement and confinement of impurities, which then tend to accumulate in the plasma core and to result in disruption or degradation of the tokamak discharge. Therefore, plasma-wall interaction, heat removal and particle exhaust will determine not only the structure and configuration of the plasma edge region, of the wall system and of the materials facing the plasma, but also the final choice of useful confinement regimes. Moreover, the potential effect of powerful {alpha}-particle heating on plasma stability and confinement has to be kept below critical values. For the latter requirement, a final answer can only be obtained in an ITER-type device where ignition and burn will become accessible. 72 refs., 12 figs.

  3. Understanding disruptions in tokamaks

    NASA Astrophysics Data System (ADS)

    Zakharov, Leonid

    2011-10-01

    Disruptions in tokamaks are known since 1963 but even now some aspects of them remain a mystery. This talk describes progress made recently in understanding disruptions. A major step forward occurred in 2007 when the importance of galvanic contact of the plasma with the wall in plasma dynamics was pointed out. The toroidal asymmetry of plasma current, observed in JET vertical disruptions, was explained by the theory of the wall touching kink mode. The currents shared by the plasma with the wall and responsible for the asymmetry were identified as generated by the kink mode. Such currents are referred to as Hiro currents. They have shown exceptional consistency with the entire JET disruption data base (more than 5500 cases) and ruled out the long lasting interpretation based on ``halo currents,'' which contradict experiments even in the sign of the measured asymmetry. Accordingly, the sideways forces are understood and their scaling from JET to ITER was justified. Hiro currents provide also a plausible explanation of the current spike at the beginning of the disruptions. The important role of the plasma edge and its interaction with the wall was revealed. Based on this new understanding of disruptions, dedicated experiments on the current spike (J-TEXT, Wuhan, China) and runaway prevention by the repetitive triggering of kink modes (T-10, AUG, Tore Supra) were motivated and are in progress. Accordingly, the need for new, adaptive grid approaches to numerical simulations of disruptions became evident. In addition to the core MHD, simulations of realistic wall geometry, disruption specific plasma edge physics, plasma-wall interaction, and energetic particles need be developed. The first results of simulations of the fast MHD regime, Hiro current generation, and slower plasma decay due to a wall touching kink mode made with the new DSC code are presented. This work is supported by US DoE contract No. DE-AC02-09-CH11466.

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

    DOE PAGES

    Pace, D. C.; Lanctot, M. J.; Jackson, G. L.; ...

    2015-09-21

    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 asmore » 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.« less

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

    SciTech Connect

    Pace, D. C.; Lanctot, M. J.; Jackson, G. L.; Sandorfi, Andy M.; Smith, S. P.; Wei, Xiangdong

    2015-09-21

    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.

  6. High Energy Particles in Tokamaks

    SciTech Connect

    White, R. B.

    2008-05-14

    This lecture covers the derivation of guiding center equations in a tokamak, orbit classification, the effect of magnetic perturbations and ripple, the interaction of particles with magnetohydrodynamic modes, including passing particle resonance, toroidal Alfven mode drive and saturation, the fishbone mode, and sawtooth stabilization.

  7. Final Technical Report: Global Field Aligned Mesh and Gyrokinetic Field Solver in a Tokamak Edge Geometry

    SciTech Connect

    Cummings, Julian C.

    2013-05-15

    This project was a collaboration between researchers at the California Institute of Technology and the University of California, Irvine to investigate the utility of a global field-aligned mesh and gyrokinetic field solver for simulations of the tokamak plasma edge region. Mesh generation software from UC Irvine was tested with specific tokamak edge magnetic geometry scenarios and the quality of the meshes and the solutions to the gyrokinetic Poisson equation were evaluated.

  8. Bootstrapped tokamak with oscillating field current drive

    SciTech Connect

    Weening, R.H. )

    1993-07-01

    A magnetic helicity conserving mean-field Ohm's law is used to study bootstrapped tokamaks with oscillating field current drive. The Ohm's law leads to the conclusion that the tokamak bootstrap effect can convert the largely alternating current of oscillating field current drive into a direct toroidal plasma current. This plasma current rectification is due to the intrinsically nonlinear nature of the tokamak bootstrap effect, and suggests that it may be possible to maintain the toroidal current of a tokamak reactor by supplementing the bootstrap current with oscillating field current drive. Steady-state tokamak fusion reactors operating with oscillating field current drive could provide an alternative to tokamak reactors operating with external current drive.

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

    NASA Astrophysics Data System (ADS)

    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.

  10. Options for an ignited tokamak

    SciTech Connect

    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.

  11. Cryogenic needs for future tokamaks

    NASA Astrophysics Data System (ADS)

    Katheder, H.

    The ITER tokamak is a machine using superconducting magnets. The windings of these magnets will be subjected to high heat loads resulting from a combination of nuclear energy absorption and AC-losses. It is estimated that about 100 kW at 4.5 K are needed. The total cooling mass flow rate will be around 10 - 15 kg/s. In addition to the large cryogenic power required for the superconducting magnets cryogenic power is also needed for refrigerated radiation shield, various cryopumps, fuel processing and test beds. A general description of the overall layout and the envisaged refrigerator cycle, necessary cold pumps and ancillary equipment is given. The basic cryogenic layout for the ITER tokakmak design, as developed during the conceptual design phase and a short overview about existing tokamak designs using superconducting magnets is given.

  12. Transport Equations In Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Callen, J. D.

    2009-11-01

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for: neoclassical effects on the parallel Ohm's law (trapped particle effects on resistivity, bootstrap current); fluctuation-induced transport; heating, current-drive and flow sources and sinks; small B field non-axisymmetries; magnetic field transients etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed recently using a kinetic-based framework. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales (and constraints they impose) are considered sequentially: compressional Alfv'en waves (Grad-Shafranov equilibrium, ion radial force balance); sound waves (pressure constant along field lines, incompressible flows within a flux surface); and ion collisions (damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on the plasma fluid: 7 ambipolar collision-based ones (classical, neoclassical, etc.) and 8 non-ambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients etc.). The plasma toroidal rotation equation [1] results from setting to zero the net radial current induced by the non-ambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the non-ambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The resultant transport equations will be presented and contrasted with the usual ones. [4pt] [1] J.D. Callen, A.J. Cole, C.C. Hegna, ``Toroidal Rotation In

  13. Magnetic island formation in tokamaks

    SciTech Connect

    Yoshikawa, S.

    1989-04-01

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

  14. Equilibrium Reconstruction in EAST Tokamak

    NASA Astrophysics Data System (ADS)

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

    2009-04-01

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

  15. Magnetic confinement experiment. I: Tokamaks

    SciTech Connect

    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.

  16. LONG PULSE ADVANCED TOKAMAK DISCHARGES IN THE DIII-D TOKAMAK

    SciTech Connect

    P.I. PETERSEN

    2002-06-01

    One of the main goals for the DIII-D research program is to establish an advanced tokamak plasma with high bootstrap current fraction that can be sustained in-principle steady-state. Substantial progress has been made in several areas during the last year. The resistive wall mode stabilization has been done with spinning plasmas in which the plasma pressure has been extended well above the no-wall beta limit. The 3/2 neoclassical tearing mode has been stabilized by the injection of ECH into the magnetic islands, which drives current to substitute the missing bootstrap current. In these experiments either the plasma was moved or the toroidal field was changed to overlap the ECCD resonance with the location of the NTMs. Effective disruption mitigation has been obtained by massive noble gas injection into shots where disruptions were deliberately triggered. The massive gas puff causes a fast and clean current quench with essentially all the plasma energy radiated fairly uniformly to the vessel walls. The run-away electrons that are normally seen accompanying disruptions are suppressed by the large density of electrons still bound on the impurity nuclei. Major elements required to establish integrated, long-pulse, advanced tokamak operations have been achieved in DIII-D: {beta}{sub T} = 4.2%, {beta}{sub p} = 2, f{sub BS} = 65%, and {beta}{sub N}H{sub 89} = 10 for 600 ms ({approx} 4{tau}{sub E}). The next challenge is to integrate the different elements, which will be the goal for the next five years when additional control will be available. Twelve resistive wall mode coils are scheduled to be installed in DIII-D during the summer of 2003. The future plans include upgrading the tokamak pulse length capability and increasing the ECH power, to control the current profile evolution.

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

    SciTech Connect

    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.

  18. Leakage of runaway electrons from tokamaks

    SciTech Connect

    Wong, K.L.

    1982-02-01

    Runaway electron orbits are calculated in a tokamak magnetic field. It is shown that these electrons tend to drift towards a larger major radius with a velocity v Vector/sub R/ = qcE/B/sub 0/ R. This effect may be relevant to some recent experimental observations in tokamaks.

  19. Task toward a Realization of Commercial Tokamak Fusion Plants in 2050 -The Role of ITER and the Succeeding Developments- 3.Fusion Plasma Research toward Fusion Power Plants

    NASA Astrophysics Data System (ADS)

    Kamada, Yutaka; Shimada, Michiya; Miura, Yukitoshi; Ogawa, Yuichi

    This section discusses fusion plasma research that needs to be carried out to develop fusion power plants. Burning plasma, in which self-heating by energetic alph aparticles plays an essential role, should be recognized as autonomous system. This is quite different from present plasma experiments, suggesting a possibility to yield some qualitative changes in fusion plasma research. Research with ITER is strongly expected to contribute to this burning plasma physics. In addition, plasma performance in steady-state and at high beta is very important in fusion power plants from the engineering and economical viewpoints. Plasma parameters expected for fusion power plants are discussed, and present status of experimental research is reviewed. Research in devices other than ITER with unique features would be instrumental for exploring high performance plasmas. A necessity of research complementary to ITER plasma is discussed.

  20. Linear optimal control of tokamak fusion devices

    SciTech Connect

    Kessel, C.E.; Firestone, M.A.; Conn, R.W.

    1989-05-01

    The control of plasma position, shape and current in a tokamak fusion reactor is examined using linear optimal control. These advanced tokamaks are characterized by non up-down symmetric coils and structure, thick structure surrounding the plasma, eddy currents, shaped plasmas, superconducting coils, vertically unstable plasmas, and hybrid function coils providing ohmic heating, vertical field, radial field, and shaping field. Models of the electromagnetic environment in a tokamak are derived and used to construct control gains that are tested in nonlinear simulations with initial perturbations. The issues of applying linear optimal control to advanced tokamaks are addressed, including complex equilibrium control, choice of cost functional weights, the coil voltage limit, discrete control, and order reduction. Results indicate that the linear optimal control is a feasible technique for controlling advanced tokamaks where the more common classical control will be severely strained or will not work. 28 refs., 13 figs.

  1. Natural current profiles in a tokamak

    SciTech Connect

    Taylor, J.B.

    1990-08-01

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

  2. Plasma diagnostics for the compact ignition tokamak

    SciTech Connect

    Medley, S.S.; Young, K.M.

    1988-06-01

    The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10/sup 21/m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs.

  3. Phase Contrast Imaging on the HL-2A Tokamak

    NASA Astrophysics Data System (ADS)

    Yu, Yi; Gong, Shaobo; Xu, Min; Jiang, Wei; Zhong, Wulv; Shi, Zhongbin; Wang, Huajie; Wu, Yifan; Yuan, Boda; Lan, Tao; Ye, Minyou; Duan, Xuru; HL-2A Team

    2016-10-01

    In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum wavenumber of 66 cm-1. The designed wavenumber resolution is 2.09cm-1, and the time resolution is higher than 0.2 μs. The broad kρs ranging from 0.34 to 13.37 makes it suitable for turbulence measurement. An upgraded PCI system is also discussed, which is designed for the HL-2M tokamak. Supported by the National Magnetic Confinement Fusion Energy Research Project (Grant No. 2015GB120002), the National Natural Science Foundation of China (Grant No. 11375053, 11105144, 10905057, 11535013).

  4. Transport equations in tokamak plasmas

    SciTech Connect

    Callen, J. D.; Hegna, C. C.; Cole, A. J.

    2010-05-15

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfven waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The 'mean field' effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The

  5. Breakdown in the pretext tokamak

    SciTech Connect

    Benesch, J.F.

    1981-06-01

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

  6. Tokamak Physics Experiment divertor design

    SciTech Connect

    Anderson, P.M.

    1995-12-31

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

  7. Predictive Modeling of Tokamak Configurations*

    NASA Astrophysics Data System (ADS)

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

    2001-10-01

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

  8. Theoretical Transport Model for Tokamaks

    NASA Astrophysics Data System (ADS)

    Ghanem, Elsayed Mohammad

    In the present thesis work a theoretical transport model is suggested to study the anomalous transport of plasma particles and energy across the axisymmetric equilibrium toroidal magnetic flux surfaces in tokamaks. The model suggests a linear combination of two transport mechanisms; drift waves, which dominate the transport in the core region, and resistive ballooning modes, which dominate the transport in the edge region. The resulting unified model has been used in a predictive transport code to simulate the plasma transport in different tokamak experiments operating in both the ohmic heating phase and the low confinement mode (L-mode). For ohmic plasma, the model was used to study the saturation of energy confinement time at high plasma density. The effect of the resistive ballooning mode as a possible cause of the saturation phenomena has been investigated together with the effect of the ion temperature gradient mode. For the low confinement mode plasmas, the study has emphasized on using the model to obtain a scaling law for the energy confinement time with the various plasma parameters compared to the scaling laws that are derived based on fitting the experimental data.

  9. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. 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. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  10. Control of Dust Inventory in Tokamaks

    SciTech Connect

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Pitcher, C. S.; Taylor, N.; Furlan, J.

    2008-09-07

    Particles with sizes ranging from 100 nm to 100 {mu}m are produced in tokamaks by the interaction of the plasma with the first wall materials and divertor. Dust has not yet been of a major concern in existing tokamaks mainly because their quantities are small and these devices are not nuclear facilities. However, in ITER and in future reactors, they could represent operational and potential safety issues. The aim of this paper is thus to describe the dust creation processes in the tokamak environment. The diagnostics and removal techniques that are needed to be implemented to measure and minimise the dust inventory are also presented. The integration of these techniques into a tokamak environment is also discussed.

  11. OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS

    SciTech Connect

    LIN-LIU,YR; STAMBAUGH,RD

    2002-11-01

    OAK A271 OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS. The dependence of the ideal ballooning {beta} limit on aspect ratio, A, and elongation {kappa} is systematically explored for nearly 100% bootstrap current driven tokamak equilibria in a wide range of the shape parameters (A = 1.2-7.0, {kappa} = 1.5-6.0 with triangularity {delta} = 0.5). The critical {beta}{sub N} is shown to be optimal at {kappa} = 3.0-4.0 for all A studied and increases as A decreases with a dependence close to A{sup -0.5}. The results obtained can be used as a theoretical basis for the choice of optimum aspect ratio and elongation of next step burning plasma tokamaks or tokamak reactors.

  12. DIII-D research operations. Annual report, October 1, 1991--September 30, 1992

    SciTech Connect

    Baker, D.

    1993-05-01

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

  13. D-D tokamak reactor studies

    SciTech Connect

    Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.

    1980-11-01

    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.

  14. Excess-density-driven snakes in tokamaks

    NASA Astrophysics Data System (ADS)

    Aydemir, A. Y.; Shaing, K. C.; Waelbroeck, F. L.

    2011-10-01

    ``Snakes'' refer to sinusoidal patterns observed on space-time plots of soft-X-ray signals in tokamak plasmas. They are generally attributed to persistent and localized density perturbations that form at a rational surface after pellet injection (Weller, JET 1987), and Parker, Alcator-C 1987), or impurity accumulation (Naujoks, ASDEX 1996, Delgado-Aparicio, C-Mod 2011). It is not clear whether all snake observations have a unique origin. A likely explanation is that material trapped inside an island driven by a temperature hole leads to the observed soft-X-ray signals (Wesson 1995). More recently, it has been suggested that they could be the result of saturated nonlinear internal kinks in low, or reversed-shear geometries (Cooper 2011). We have started an examination of some of these issues using ideas from neoclassical transport theory (Shaing 2007) in conjunction with various magnetohydrodynamic models. In a RMHD model, we demonstrated that excess-density-driven bootstrap current can stabilize a resistive m = 1 island at a small amplitude, leaving a radially and poloidally localized snake-like structure. Extension of this work to more sophisticated models that include diamagnetic effects, and possibly more realistic geometries, will be presented. This research was supported by the Office of Fusion Energy Science of the U.S. Department of Energy.

  15. IMPROVED CCD DETECTORS FOR HIGH SPEED, CHARGE EXCHANGE SPECTROSCOPY STUDIES ON THE DIII-D TOKAMAK

    SciTech Connect

    K.H. BURRELL; P. GOHIL; R.J. GROEBNER; D.H. KAPLAN; D.G. NILSON; J.I. ROBINSON; D.M. THOMAS

    2002-06-01

    The tokamak plasmas utilized in magnetic fusion research provide access to some of the longest duration, highest temperature plasmas on earth. For example, on the DIII-D tokamak, plasmas lasting up to 10 seconds have been produced with ion temperatures up to 27 keV and electron temperatures up to 15 keV in toroidal plasmas with 1.7 m major radius, 0.65 m half width and 2.5 m vertical height. Typical particle densities in these plasmas are in the range of 2.0 x 10{sup 19} m{sup -3} through 2.0 x 10{sup 20} m{sup -3}. Spectroscopic measurements of line radiation from highly ionized atoms provide one of the key techniques for determining the plasma parameters in tokamak discharges.

  16. Development of Numerical Methods to Estimate the Ohmic Breakdown Scenarios of a Tokamak

    NASA Astrophysics Data System (ADS)

    Yoo, Min-Gu; Kim, Jayhyun; An, Younghwa; Hwang, Yong-Seok; Shim, Seung Bo; Lee, Hae June; Na, Yong-Su

    2011-10-01

    The ohmic breakdown is a fundamental method to initiate the plasma in a tokamak. For the robust breakdown, ohmic breakdown scenarios have to be carefully designed by optimizing the magnetic field configurations to minimize the stray magnetic fields. This research focuses on development of numerical methods to estimate the ohmic breakdown scenarios by precise analysis of the magnetic field configurations. This is essential for the robust and optimal breakdown and start-up of fusion devices especially for ITER and its beyond equipped with low toroidal electric field (ET <= 0.3 V/m). A field-line-following analysis code based on the Townsend avalanche theory and a particle simulation code are developed to analyze the breakdown characteristics of actual complex magnetic field configurations including the stray magnetic fields in tokamaks. They are applied to the ohmic breakdown scenarios of tokamaks such as KSTAR and VEST and compared with experiments.

  17. Modelling of VUV/XUV spectra from the JET Tokamak

    NASA Astrophysics Data System (ADS)

    Lawson, K. D.; Aggarwal, K. M.; Coffey, I. H.; Keenan, F. P.; Reid, R. H. G.; Zacks, J.; Contributors, JET-EFDA

    2012-05-01

    The VUV/XUV spectral region is particularly rich in lines emitted by plasmas with temperatures ≤ 30 keV used in fusion research. Examples are presented of analyses of JET tokamak data involving C VUV/XUV radiation. In the first, the C IV divertor emission is modelled, with agreement between theory and experiment to within the measurement accuracy of ~±10%. A second deals with C emission from the Scrape-off Layer (SOL) within the main chamber for which inconsistencies are found.

  18. Plasma Physics Regimes in Tokamaks with Li Walls

    SciTech Connect

    L.E. Zakharo; N.N. Gorelenkov; R.B. White; S.I. Krasheninnikov; G.V. Pereverzev

    2003-08-21

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

  19. Public Data Set: Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup

    SciTech Connect

    Hinson, Edward T.; Barr, Jayson L.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.; Perry, Justin M.

    2016-05-31

    This data set contains openly-documented, machine readable digital research data corresponding to figures published in E.T. Hinson et al., 'Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup,' Physics of Plasmas 23, 052515 (2016).

  20. Report of planning workshop on fluctuations and anomalous transport in tokamaks

    SciTech Connect

    Not Available

    1982-10-01

    The workshop was divided into three sections: review of experiments on fluctuations in tokamaks, review of theories of anomalous transport, and discussion of directions for future research and experimental/theoretical collaboration. Each session was assigned a recording secretary to take notes which were used in preparing this report. The report includes the activities, conclusions, and recommendations of the workshop.

  1. Toroidal Flow in Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Cole, A. J.; Hegna, C. C.

    2007-11-01

    Many effects influence toroidal flow evolution in tokamak plasmas. Momentum sources and radial diffusion due to axisymmetric neoclassical, paleoclassical and anomalous transport are usually considered. In addition, the toroidal flow can be affected by field errors. Small, non-axisymmetric field errors arise from coil irregularities, active control coils and collective plasma magnetic distortions (e.g., NTMs, RWMs). Resonant field errors cause localized electromagnetic torques near rational surfaces in the plasma, which can lock the plasma to the wall leading to magnetic islands and reduced confinement or disruptions. Their penetration into the plasma is limited by flow-shielding effects; but they can be amplified by the plasma response at high beta. Non-resonant field errors cause magnetic pumping and radial banana drifts, and lead to toroidal flow damping over the entire plasma. Many of these processes can also produce momentum pinch and intrinsic flow effects. This poster will seek to present a coherent picture of all these effects and suggest ways they could be tested and distinguished experimentally.

  2. Dust in tokamaks: An overview of the physical model of the dust in tokamaks code

    NASA Astrophysics Data System (ADS)

    Bacharis, Minas; Coppins, Michael; Allen, John E.

    2010-04-01

    The dynamical behavior of dust produced in tokamaks is an important issue for fusion. In this work, the current status of the dust in tokamaks (DTOKS) [J. D. Martin et al., Europhys Lett. 83, 65001 (2008)] dust transport code will be presented. A detailed description of the various elements of its underlying physical model will be given together with representative simulation results for the mega amp spherical tokamak (MAST) [A. Sykes et al., Nucl. Fusion 41, 1423 (2001)]. Furthermore, a brief description of the various components of the dust transport (DUSTT) [R. D. Smirnov et al., Plasma Phys. Controlled Fusion 49, 347 (2007)] code will also be presented in comparison with DTOKS.

  3. Activation analysis of the compact ignition tokamak

    SciTech Connect

    Selcow, E.C.

    1986-01-01

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

  4. Helicity content and tokamak applications of helicity

    SciTech Connect

    Boozer, A.H.

    1986-05-01

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

  5. Tokamak power systems studies, FY 1985

    SciTech Connect

    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.

  6. Tokamak Spectroscopy for X-Ray Astronomy

    NASA Technical Reports Server (NTRS)

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

    2000-01-01

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

  7. MHD Effects of a Ferritic Wall on Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Hughes, Paul E.

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

  8. KORC: A Kinetic Orbit Runaway Electrons code for tokamak disruptions

    NASA Astrophysics Data System (ADS)

    Carbajal Gomez, Leopoldo; Del-Castillo-Negrete, Diego; Spong, Donald; Seal, Sudip; Baylor, Larry

    2016-10-01

    Runaway electrons (RE) resulting from the violent termination of tokamak plasmas pose a serious threat to ITER due to the very high energies they can reach and deposit on the plasma facing components. Most of the current modelling of RE in fusion tokamak plasmas rely on reduced models such as the bounce-average and the test particle equations. In some scenarios, the radiation losses in these models might lead to uncertainties in the RE parameters that determine their confinement and energy limit. In order to study this in detail we have developed a new Kinetic Orbit Runaway electrons Code (KORC). KORC follows the dynamics of ensembles of relativistic electrons in the 6D phase space fully resolving gyro-motion under the influence of the Lorentz force, the Landau-Lifshiftz consistent formulation of the Abraham-Lorentz-Dirac force for radiation damping, and collisions with impurities and the background plasma. KORC is parallelized using open MP/MPI, and benefits from a modified relativistic leap-frog method along with an operator splitting scheme for solving the RE dynamics in different magnetic fields. The code is robust, conservative, and shows nearly linear strong scaling. Research sponsored by the Laboratory Directed Research and Development Program of Oak Ridge National Laboratory, managed by UT-Battelle, LLC, for the U. S. Department of Energy.

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

    NASA Astrophysics Data System (ADS)

    Burenko, Oleg

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

  10. Transformer Recharging with Alpha Channeling in Tokamaks

    SciTech Connect

    N.J. Fisch

    2009-12-21

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

  11. Analysis of sawtooth relaxation oscillations in tokamaks

    SciTech Connect

    Yamazaki, K.; McGuire, K.; Okabayashi, M.

    1982-07-01

    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.

  12. Stabilization of tokamak plasma by lithium streams

    SciTech Connect

    L.E. Zakharov

    2000-08-07

    The stabilization theory of free-boundary magnetohydrodynamic instabilities in tokamaks by liquid lithium streams driven by magnetic propulsion is formulated. While the conventional, wall-locked, resistive wall mode can be well suppressed by the flow, a new, stream-locked mode determines the limits of the flow stabilization.

  13. Simulation of runaway electrons in tokamak

    NASA Astrophysics Data System (ADS)

    Guo, Zehua; Tang, Xianzhu; McDevitt, Chris

    2015-11-01

    Runaway electrons with relativisitc energy (>Mev) are generated in tokamaks when the acceleration by parallel electric field exceeds the drag due to Coulomb collisions with the bulk plasma. Carrying about 70% of the ITER thermal current (15MA), they can possibly cause severe damage to tokamak facing components. Here we report the development of a solver for computing the evolution of runaway electron distribution in tokamak geometries. Essential effects from Coulomb collisions, radiation losses, toroidal effects and the radial transport are included on the same footings. Numerical techniques (implicit-explicit time-stepping, KT/NT central schemes) to overcome the difficulties arising from the wide spread of time scales in runaway electron dynamics and the hyperbolic nature of the relativistic Fokker-Planck equation will be discussed. We will use the solver to study two important physics: 1) the presence of stable point in the phase space and its relation to the electric field threshold; 2) the radial transport of runaways in tokamak geometry and its effects on the distribution function. Work supported by DOE via LANL-LDRD.

  14. Elementary Processes Underlying Alpha Channeling in Tokamaks

    SciTech Connect

    NM.J. Fisch

    2012-06-15

    Alpha channeling in tokamaks is speculative, but also extraordinarily attractive. Waves that can accomplish this effect have been identified. Key aspects of the theory now enjoy experimental confirmation. This paper will review the elementary processes of wave-particle interactions in plasma that underlie the alpha channeling effect

  15. Diagnostics for neutral-beam-heated tokamaks

    SciTech Connect

    Goldston, R.J.

    1982-12-01

    Diagnostic techniques for neutral-beam-heated tokamak plasmas fall into three categories: (1) magnetic diagnostics for measurements of gross stored energy, (2) profile diagnostics for measurements of stored thermal and beam energy, impurity content and plasma rotation, and (3) fast time resolution diagnostics to study MHD fluctuations and micro-turbulence.

  16. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    SciTech Connect

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

  17. DIAMOND WIRE CUTTING OF THE TOKAMAK FUSION TEST REACTOR

    SciTech Connect

    Rule, Keith; Perry, Erik; Parsells, Robert

    2003-02-27

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the techno logy was improved and redesigned for the actual cutting of the vacuum vessel. 10 complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D activity.

  18. Advanced tokamak scenario developments for the next step

    NASA Astrophysics Data System (ADS)

    Joffrin, E.

    2007-12-01

    The objective of advanced tokamak scenario research is to provide a candidate plasma scenario for continuous operation in a fusion power plant. The optimization of the self-generated non-inductive current by the bootstrap mechanism up to a level of 50% and above using high plasma pressure and improved confinement are the necessary conditions to achieve this goal. The two main candidate scenarios for continuous operation, the steady state scenario and long duration (up to 3000 s) high neutron fluency scenario (the hybrid scenario), both face physics challenges in terms of confinement, stability, power exhaust and plasma control. Resistive wall modes and Alfvénic fast ion driven instabilities are the main limitations for operating the steady state scenario at high pressure and low magnetic shear. In addition, this scenario demands a high degree of control over the plasma current and pressure profile and the steady state heat load on in-vessel plasma facing components. Understanding the confinement properties of hybrid scenario is still an outstanding issue as well as its modelling for ITER in particular with regard to the H-mode pedestal parameters. This scenario will also require active current profile control, although, less demanding than for the steady state scenario. To operate advanced tokamak scenario, broad current and pressure profile control appears as a necessary requirement on ITER actuators, in addition to the tools required for instability control such as error field coils or electron cyclotron current drive.

  19. Sensitivity of magnetic field-line pitch angle measurements to sawtooth events in tokamaks

    NASA Astrophysics Data System (ADS)

    Ko, J.

    2016-11-01

    The sensitivity of the pitch angle profiles measured by the motional Stark effect (MSE) diagnostic to the evolution of the safety factor, q, profiles during the tokamak sawtooth events has been investigated for Korea Superconducting Tokamak Advanced Research (KSTAR). An analytic relation between the tokamak pitch angle, γ, and q estimates that Δγ ˜ 0.1° is required for detecting Δq ˜ 0.05 near the magnetic axis (not at the magnetic axis, though). The pitch angle becomes less sensitive to the same Δq for the middle and outer regions of the plasma (Δγ ˜ 0.5°). At the magnetic axis, it is not straightforward to directly relate the γ sensitivity to Δq since the gradient of γ(R), where R is the major radius of the tokamak, is involved. Many of the MSE data obtained from the 2015 KSTAR campaign, when calibrated carefully, can meet these requirements with the time integration down to 10 ms. The analysis with the measured data shows that the pitch angle profiles and their gradients near the magnetic axis can resolve the change of the q profiles including the central safety factor, q0, during the sawtooth events.

  20. Nonlinear Effects at Tokamak Electron Cyclotron Resonance in Inhomogeneous Magnetic Field.*

    NASA Astrophysics Data System (ADS)

    Stefan, V.

    1996-11-01

    Nonlinear interaction of X- and O- modes with drift plasma waves is studied. The drift waves with frequency given by ωD ~ Ωc (ρ_e/r)^2 (ρe electron Larmor radius, Ωe electron cyclotron frequency, r small tokamak radius, where nabla Ω / Ωe ~ 1/R (for large tokamaks R ~ r)), are coupled to driver pump via scattering instability. Nonlocality of the interaction is taken into account. It is shown that nonlinear mechanism of interaction (Brillouin scattering) can be used as a tool for dynamic rf confinement^1 of tokamak plasmas. Particularly, it is possible to achieve longer confinement times due to suppression of drift wave turbulence. Supported by Tesla Laboratories, Inc., La Jolla, CA 92038-2946. ^1M.N. Rosenbluth (Editor-in-Chief). New Ideas in Tokamak Confinement. Research Trends in Physics Series of the La Jolla International School of Physics, The Institute for Advanced Physics Studies, La Jolla, CA (AIP Press, New York, 1994).

  1. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks. Annual report, November 16, 1991--November 15, 1992

    SciTech Connect

    Scharer, J.E.

    1992-12-31

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  2. Equilibrium system analysis in a tokamak ignition experiment

    SciTech Connect

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

    1989-10-01

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

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

    SciTech Connect

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

    1989-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2009-05-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

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

    SciTech Connect

    BURRELL,KH

    2002-11-01

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

  6. Neural net prediction of tokamak plasma disruptions

    NASA Astrophysics Data System (ADS)

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

    1994-10-01

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

  7. Rapidly Moving Divertor Plates In A Tokamak

    SciTech Connect

    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.

  8. Global migration of impurities in tokamaks

    NASA Astrophysics Data System (ADS)

    Hakola, A.; Airila, M. I.; Björkas, C.; Borodin, D.; Brezinsek, S.; Coad, J. P.; Groth, M.; Järvinen, A.; Kirschner, A.; Koivuranta, S.; Krieger, K.; Kurki-Suonio, T.; Likonen, J.; Lindholm, V.; Makkonen, T.; Mayer, M.; Miettunen, J.; Müller, H. W.; Neu, R.; Petersson, P.; Rohde, V.; Rubel, M.; Widdowson, A.; the ASDEX Upgrade Team; Contributors, JET-EFDA

    2013-12-01

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

  9. Filamentary probe on the COMPASS tokamak

    NASA Astrophysics Data System (ADS)

    Kovarik, K.; Duran, I.; Stockel, J.; Seidl, J.; Adamek, J.; Spolaore, M.; Vianello, N.; Hacek, P.; Hron, M.; Panek, R.

    2017-03-01

    This paper describes a new filamentary probe recently introduced on the COMPASS tokamak. It allows the measurement of electrostatic and magnetic properties of the filaments and their changes in dependence on distance from the separatrix in the region between a divertor and midplane. The probe head is mounted on a manipulator moving the probe radially on a shot-to-shot basis. This configuration is suitable for the long term statistical measurement of the plasma filaments and the measurement of their evolution during their propagation from the separatrix to the wall. The basics of the filamentary probe construction, the evolution of the plasma parameters, and first conditional averages of the plasma filaments in the scrape-off layer of the COMPASS tokamak during the L-mode regime are presented.

  10. Boundary Plasma Turbulence Simulations for Tokamaks

    SciTech Connect

    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.

  11. Tokamak with liquid metal toroidal field coil

    DOEpatents

    Ohkawa, Tihiro; Schaffer, Michael J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof.

  12. Confinement scaling and ignition in tokamaks

    SciTech Connect

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

    1985-10-01

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

  13. Self-Organized Stationary States of Tokamaks

    SciTech Connect

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

    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."

  14. Plasma filamentation in the Rijnhuizen tokamak RTP

    SciTech Connect

    Lopes Cardozo, N.J.; Schueller, F.C.; Barth, C.J.; Chu, C.C.; Pijper, F.J.; Lok, J.; Oomens, A.A.M. )

    1994-07-11

    Evidence for small scale magnetic structures in the Rijnhuizen tokamak RTP is presented. These are manifest through steps and peaks in the electron temperature and pressure, measured with multiposition Thomson scattering. During central electron cyclotron heating, several filaments of high pressure are found in the power deposition region. They live hundreds of microseconds. Near the sawtooth inversion radius a step'' in the temperature profile occurs. Further out, quasiperiodic structures are observed, in both Ohmic and heated discharges.

  15. Self-Organized Stationary States of Tokamaks.

    PubMed

    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."

  16. Neutral-beam current drive in tokamaks

    SciTech Connect

    Devoto, R.S.

    1986-01-01

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

  17. Forced Magnetic Reconnection In A Tokamak Plasma

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Hegna, C. C.

    2015-11-01

    The theory of forced magnetic field reconnection induced by an externally imposed resonant magnetic perturbation usually uses a sheared slab or cylindrical magnetic field model and often focuses on the potential time-asymptotic induced magnetic island state. However, tokamak plasmas have significant magnetic geometry and dynamical plasma toroidal rotation screening effects. Also, finite ion Larmor radius (FLR) and banana width (FBW) effects can damp and thus limit the width of a nascent magnetic island. A theory that is more applicable for tokamak plasmas is being developed. This new model of the dynamics of forced magnetic reconnection considers a single helicity magnetic perturbation in the tokamak magnetic field geometry, uses a kinetically-derived collisional parallel electron flow response, and employs a comprehensive dynamical equation for the plasma toroidal rotation frequency. It is being used to explore the dynamics of bifurcation into a magnetically reconnected state in the thin singular layer around the rational surface, evolution into a generalized Rutherford regime where the island width exceeds the singular layer width, and assess the island width limiting effects of FLR and FBW polarization currents. Support by DoE grants DE-FG02-86ER53218, DE-FG02-92ER54139.

  18. Spherical tokamaks with plasma centre-post

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2013-10-01

    The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.

  19. Interactive, multiobjective Bayesian optimization of tokamak scenarios

    NASA Astrophysics Data System (ADS)

    Urban, Jakub; Artaud, Jean-François

    2016-10-01

    Bayesian optimization is applied to tokamak scenario optimizations. The key advantages are 1) a reduced number of objective function evaluations, 2) no need for derivatives, and 3) the possibility to include a prior knowledge. This is of a great value for optimizing tokamak scenarios, where several (competing) objectives with often unknown magnitudes exist and the number of parameters is large (>10). The first two properties imply that Bayesian optimization is well suited for heavy, complex objective functions. Reusing previous iterations as priors for next optimization steps effectively enables interactive, multiobjective optimizations, regardless of whether a human decision maker is included or not. We show that these features make Bayesian optimization an outstanding tool for optimizing tokamak scenarios. Objective functions and constraints, targeting, e.g., fusion gain, flux consumption, coils currents limits or q-profile, can be assembled interactively. The optimized parameter vector may include actuators like plasma current or heating waveforms. We demonstrate the capabilities on optimizing ITER and DEMO-like scenarios, simulated by the METIS code.

  20. The Spherical Tokamak MEDUSA for Mexico

    NASA Astrophysics Data System (ADS)

    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 < 0.14m, a < 0.10m, BT < 0.5T, Ip < 40kA, 3ms pulse) is currently being recomissioned at the Universidad Autónoma de Nuevo León, Mexico, as part of an agreement between the Faculties of Mech.-Elect. Eng. and Phy. Sci.-Maths. The main objective for having MEDUSA is to train students in plasma physics & technical related issues, aiming a full design of a medium size device (e.g. Tokamak-T). Details of technical modifications and a preliminary scientific programme will be presented. MEDUSA-MX will also benefit any developments in the existing Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

  1. Nondiffusive plasma transport at tokamak edge

    NASA Astrophysics Data System (ADS)

    Krasheninnikov, S. I.

    2000-10-01

    Recent findings show that cross field edge plasma transport at tokamak edge does not necessarily obey a simple diffusive law [1], the only type of a transport model applied so far in the macroscopic modeling of edge plasma transport. Cross field edge transport is more likely due to plasma filamentation with a ballistic motion of the filaments towards the first wall. Moreover, it so fast that plasma recycles on the main chamber first wall rather than to flow into divertor as conventional picture of edge plasma fluxes suggests. Crudely speaking particle recycling wise diverted tokamak operates in a limiter regime due to fast anomalous non-diffusive cross field plasma transport. Obviously that this newly found feature of edge plasma anomalous transport can significantly alter a design of any future reactor relevant tokamaks. Here we present a simple model describing the motion of the filaments in the scrape off layer and discuss it implications for experimental observations. [1] M. Umansky, S. I. Krasheninnikov, B. LaBombard, B. Lipschultz, and J. L. Terry, Phys. Plasmas 6 (1999) 2791; M. Umansky, S. I. Krasheninnikov, B. LaBombard and J. L. Terry, Phys. Plasmas 5 (1998) 3373.

  2. Edge-localized-modes in tokamaks

    SciTech Connect

    Leonard, A. W.

    2014-09-15

    Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively, rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heat flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. Encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.

  3. Remote feedback stabilization of tokamak instabilities

    SciTech Connect

    Sen, A.K. )

    1994-05-01

    A novel remote suppressor consisting of an injected ion beam has been used for the stabilization of plasma instabilities. A collisionless curvature-driven trapped-particle instability, an [bold E][times][bold B] flute mode and an ion temperature gradient (ITG) instability have been successfully suppressed down to noise levels using this scheme. Furthermore, the first experimental demonstration of a multimode feedback stabilization with a single sensor--suppressor pair has been achieved. Two modes (an [bold E][times][bold B] flute and an ITG mode) were simultaneously stabilized with a simple state-feedback-type method where more state'' information was generated from a single-sensor Langmuir probe by appropriate signal processing. The above experiments may be considered as paradigms for controlling several important tokamak instabilities. First, feedback suppression of edge fluctuations in a tokamak with a suitable form of insulated segmented poloidal limiter sections used as Langmuir-probe-like suppressors is proposed. Other feedback control schemes are proposed for the suppression of electrostatic core fluctuations via appropriately phased ion density input from a modulated neutral beam. Most importantly, a scheme to control major disruptions in tokamaks via feedback suppression of kink (and possibly) tearing modes is discussed. This may be accomplished by using a modulated neutral beam suppressor in a feedback loop, which will supply a momentum input of appropriate phase and amplitude. Simple theoretical models predict modest levels of beam energy, current, and power.

  4. Composition And Electrical Properties Of Dust From Tokamak Compass

    SciTech Connect

    Vaverka, J.; Beranek, M.; Pavlu, J.; Richterova, I.; Vysinka, M.; Safrankova, J.; Nemecek, Z.

    2011-11-29

    In spite of the fact that fusion is a subject of the study for many years, there are still a lot of open questions. One of the interesting topics in fusion research is a presence of dust grains in reactors. In the paper, dust grains born in tokamak Compass are studied and compared with samples of a spherical geometry and well known composition. A unique experimental setup was used for investigations of charging properties of such grains and the SEM and EDX spectroscopy was applied for a study of grain composition. We focus on the secondary emission because this process plays a prominent role when a portion of energetic electrons is present in surroundings of a particular grain. It was shown that depending on the grain size and material, energetic electrons charge the grains to positive potentials comparable with the energy of impinging electrons.

  5. A divertor plasma configuration design method for tokamaks

    NASA Astrophysics Data System (ADS)

    Guo, Yong; Xiao, Bing-Jia; Liu, Lei; Yang, Fei; Wang, Yuehang; Qiu, Qinglai

    2016-11-01

    The efficient and safe operation of large fusion devices strongly relies on the plasma configuration inside the vacuum chamber. It is important to construct the proper plasma equilibrium with a desired plasma configuration. In order to construct the target configuration, a shape constraint module has been developed in the tokamak simulation code (TSC), which controls the poloidal flux and the magnetic field at several defined control points. It is used to construct the double null, lower single null, and quasi-snowflake configurations for the required target shape and calculate the required PF coils current. The flexibility and practicability of this method have been verified by the simulated results. Project supported by the National Magnetic Confinement Fusion Research Program of China (Grant Nos. 2014GB103000 and 2014GB110003), the National Natural Science Foundation of China (Grant Nos. 11305216, 11305209, and 11375191), and External Cooperation Program of BIC, Chinese Academy of Sciences (Grant No. GJHZ201303).

  6. Tokamak power exhaust with the snowflake divertor: Present results and outstanding issues

    SciTech Connect

    Soukhanovskii, V. A.; Xu, X.

    2015-09-15

    Here, a snowflake divertor magnetic configuration (Ryutov in Phys Plasmas 14(6):064502, 2007) with the second-order poloidal field null offers a number of possible advantages for tokamak plasma heat and particle exhaust in comparison with the standard poloidal divertor with the first-order null. Results from snowflake divertor experiments are briefly reviewed and future directions for research in this area are outlined.

  7. Public Data Set: High Confinement Mode and Edge Localized Mode Characteristics in a Near-Unity Aspect Ratio Tokamak

    SciTech Connect

    Thome, Kathreen E.; Bongard, Michael W.; Barr, Jayson L.; Bodner, Grant M.; Burke, Marcus G.; Fonck, Raymond J.; Kriete, David M.; Perry, Justin M.; Schlossberg, David J.

    2016-04-27

    This data set contains openly-documented, machine readable digital research data corresponding to figures published in K.E. Thome et al., 'High Confinement Mode and Edge Localized Mode Characteristics in a Near-Unity Aspect Ratio Tokamak,' Phys. Rev. Lett. 116, 175001 (2016).

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

    SciTech Connect

    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.

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

    SciTech Connect

    Killeen, J.; Mirin, A.A.; McCoy, M.G.

    1980-01-01

    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.

  10. Advanced tokamak operating modes in TPX and ITER

    SciTech Connect

    Nevins, W.M.

    1994-12-31

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

  11. A simulation study of a controlled tokamak plasma

    NASA Astrophysics Data System (ADS)

    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.

  12. Effects of neoclassical toroidal viscosity induced by the intrinsic error fields and toroidal field ripple on the toroidal rotation in tokamaks

    NASA Astrophysics Data System (ADS)

    Lee, H. H.; Seol, J.; Ko, W. H.; Terzolo, L.; Aydemir, A. Y.; In, Y.; Ghim, Y.-c.; Lee, S. G.

    2016-08-01

    Effects of neoclassical toroidal viscosity (NTV) induced by intrinsic error fields and toroidal field ripple on cocurrent toroidal rotation in H-mode tokamak plasmas are investigated. It is expected that large NTV torque can be localized at the edge region through the 1/ν-regime in the vicinity of E r ˜ 0 in the cocurrent rotating H-mode plasma. Numerical simulation on toroidal rotation demonstrates that the edge localized NTV torque determined by the intrinsic error fields and toroidal field ripples in the level of most tokamaks can damp the toroidal rotation velocity over the whole region while reducing the toroidal rotation pedestal which is clearly observed in Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak. It is found that the NTV torque changes the toroidal rotation gradient in the pedestal region dramatically, but the toroidal rotation profile in the core region responds rigidly without a change in the gradient. On the other hand, it shows that the NTV torque induced by the intrinsic error fields and toroidal field ripple in the level of the KSTAR tokamak, which are expected to be smaller than most tokamaks by at least one order of magnitude, is negligible in determining the toroidal rotation velocity profile. Experimental observation on the toroidal rotation change by the externally applied nonaxisymmetric magnetic fields on KSTAR also suggests that NTV torque arising from nonaxisymmetric magnetic fields can damp the toroidal rotation over the whole region while diminishing the toroidal rotation pedestal.

  13. Hybrid Fusion: The Only Viable Development Path for Tokamaks?

    NASA Astrophysics Data System (ADS)

    Manheimer, Wallace

    2009-03-01

    The world needs a great deal of carbon free energy, and soon, for civilization to continue. Fusion's goal is to develop such a carbon free energy source. For the last 4 decades, tokamaks have been the best magnetic fusion has to offer. But what if its development stops short of commercial fusion? This paper introduces `conservative design principles' for tokamaks. These are very simple, are reasonably based in theory, and have always constrained tokamak operation. Assuming they continue to do so, it is unlikely that tokamaks will ever make it as commercial reactors. This is independent of their confinement properties. However because of the large additional gain in hybrid fusion, tokamaks reactors look like they can make it as hybrid fuel producers, and provide large scale power by mid century or shortly thereafter.

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

    SciTech Connect

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

    1995-04-01

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

  15. First neutral beam injection experiments on KSTAR tokamak.

    PubMed

    Jeong, S H; Chang, D H; Kim, T S; In, S R; Lee, K W; Jin, J T; Chang, D S; Oh, B H; Bae, Y S; Kim, J S; Park, H T; Watanabe, K; Inoue, T; Kashiwagi, M; Dairaku, M; Tobari, H; Hanada, M

    2012-02-01

    The first neutral beam (NB) injection system of the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak was partially completed in 2010 with only 1∕3 of its full design capability, and NB heating experiments were carried out during the 2010 KSTAR operation campaign. The ion source is composed of a JAEA bucket plasma generator and a KAERI large multi-aperture accelerator assembly, which is designed to deliver a 1.5 MW, NB power of deuterium at 95 keV. Before the beam injection experiments, discharge, and beam extraction characteristics of the ion source were investigated. The ion source has good beam optics in a broad range of beam perveance. The optimum perveance is 1.1-1.3 μP, and the minimum beam divergence angle measured by the Doppler shift spectroscopy is 0.8°. The ion species ratio is D(+):D(2)(+):D(3)(+) = 75:20:5 at beam current density of 85 mA/cm(2). The arc efficiency is more than 1.0 A∕kW. In the 2010 KSTAR campaign, a deuterium NB power of 0.7-1.5 MW was successfully injected into the KSTAR plasma with a beam energy of 70-90 keV. L-H transitions were observed within a wide range of beam powers relative to a threshold value. The edge pedestal formation in the T(i) and T(e) profiles was verified through CES and electron cyclotron emission diagnostics. In every deuterium NB injection, a burst of D-D neutrons was recorded, and increases in the ion temperature and plasma stored energy were found.

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

    SciTech Connect

    Chektybayev, B. Shapovalov, G.; Kolodeshnikov, A.

    2015-05-15

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

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

    PubMed

    Chektybayev, B; Shapovalov, G; Kolodeshnikov, A

    2015-05-01

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

  18. Researchers build a secure plasma prison

    SciTech Connect

    Glanz, J.

    1995-07-28

    Research groups at Princeton University`s Tokamak Fusion Test Reactor and at the DIII-D tokamak at General Atomics in San Diego have made a major breakthrough. By tailoring the magnetic fields with unprecedented finesse, they appear to have tamed the plasma instabilities that rattle and tear the fragile magnetic cage, allowing particles to leak out and limiting a tokamak`s performance. In the process they increased the central density of the plasma by as much as threefold and reduced the particle leakage by a factor of 50.

  19. Mathematical modeling plasma transport in tokamaks

    SciTech Connect

    Quiang, Ji

    1997-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

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

  1. Largescale Long-term particle Simulations of Runaway electrons in Tokamaks

    NASA Astrophysics Data System (ADS)

    Liu, Jian; Qin, Hong; Wang, Yulei

    2016-10-01

    To understand runaway dynamical behavior is crucial to assess the safety of tokamaks. Though many important analytical and numerical results have been achieved, the overall dynamic behaviors of runaway electrons in a realistic tokamak configuration is still rather vague. In this work, the secular full-orbit simulations of runaway electrons are carried out based on a relativistic volume-preserving algorithm. Detailed phase-space behaviors of runaway electrons are investigated in different timescales spanning 11 orders. A detailed analysis of the collisionless neoclassical scattering is provided when considering the coupling between the rotation of momentum vector and the background field. In large timescale, the initial condition of runaway electrons in phase space globally influences the runaway distribution. It is discovered that parameters and field configuration of tokamaks can modify the runaway electron dynamics significantly. Simulations on 10 million cores of supercomputer using the APT code have been completed. A resolution of 107 in phase space is used, and simulations are performed for 1011 time steps. Largescale simulations show that in a realistic fusion reactor, the concern of runaway electrons is not as serious as previously thought. This research was supported by National Magnetic Connement Fusion Energy Research Project (2015GB111003, 2014GB124005), the National Natural Science Foundation of China (NSFC-11575185, 11575186) and the GeoAlgorithmic Plasma Simulator (GAPS) Project.

  2. Electrostatic analysis of the tokamak edge plasma

    SciTech Connect

    Motley, R.W.

    1981-07-01

    The intrusion of an equipotential poloidal limiter into the edge plasma of a circular tokamak discharge distorts the axisymmetry in two ways: (1) it (partially) shorts out the top-to-bottom Pfirsch-Schlueter driving potentials, and (2) it creates zones of back current flow into the limiter. The resulting boundary mismatch between the outer layers and the inner axisymmetric Pfirsch-Schlueter layer provides free energy to drive the edge plasma unstable. Special limiters are proposed to symmetrize the edge plasma and thereby reduce the electrical and MHD activity in the boundary layer.

  3. Tokamak equilibria with reversed current density.

    PubMed

    Martynov, A A; Medvedev, S Yu; Villard, L

    2003-08-22

    Observations of nearly zero toroidal current in the central region of tokamaks (the "current hole") raises the question of the existence of toroidal equilibria with very low or reversed current in the core. The solutions of the Grad-Shafranov equilibrium equation with hollow toroidal current density profile including negative current density in the plasma center are investigated. Solutions of the corresponding eigenvalue problem provide simple examples of such equilibrium configurations. More realistic equilibria with toroidal current density reversal are computed using a new equilibrium problem formulation and computational algorithm which do not assume nested magnetic surfaces.

  4. Viscosity in the edge of tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Stacey, W. M.

    1993-05-01

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

  5. Self-Organized Stationary States of Tokamaks

    SciTech Connect

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

    2015-11-17

    We demonstrate that in a 3D resistive magnetohydrodynamic (MHD) 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 non-linearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary non-sawtoothing “hybrid” discharges, often referred to as “flux-pumping”.

  6. Viscosity in the edge of tokamak plasmas

    SciTech Connect

    Stacey, W.M.

    1993-05-01

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

  7. Nonlinear gyrokinetic equations for tokamak microturbulence

    SciTech Connect

    Hahm, T.S.

    1988-05-01

    A nonlinear electrostatic gyrokinetic Vlasov equation, as well as Poisson equation, has been derived in a form suitable for particle simulation studies of tokamak microturbulence and associated anomalous transport. This work differs from the existing nonlinear gyrokinetic theories in toroidal geometry, since the present equations conserve energy while retaining the crucial linear and nonlinear polarization physics. In the derivation, the action-variational Lie perturbation method is utilized in order to preserve the Hamiltonian structure of the original Vlasov-Poisson system. Emphasis is placed on the dominant physics of the collective fluctuations in toroidal geometry, rather than on details of particle orbits. 13 refs.

  8. An efficient transport solver for tokamak plasmas

    DOE PAGES

    Park, Jin Myung; Murakami, Masanori; St. John, H. E.; ...

    2017-01-03

    A simple approach to efficiently solve a coupled set of 1-D diffusion-type transport equations with a stiff transport model for tokamak plasmas is presented based on the 4th order accurate Interpolated Differential Operator scheme along with a nonlinear iteration method derived from a root-finding algorithm. Here, numerical tests using the Trapped Gyro-Landau-Fluid model show that the presented high order method provides an accurate transport solution using a small number of grid points with robust nonlinear convergence.

  9. Midplane Faraday rotation: A tokamak densitometer

    NASA Astrophysics Data System (ADS)

    Jobes, F. C.

    1995-01-01

    The density in a tokamak can be determined by measuring the Faraday rotation of a laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then ne(R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, such as ITER, TPX, or JT-60, a 10.6 μm laser would be appropriate. If the machine operated at a full field of 10-50 T m and a peak density of 2.5×1020/m3, the rotation angle would be quite large—about 15°-75° per pass. An elegant measurement system can be made up from a single laser beam diffracted off a moving grating to form a fan of ˜10 probe beams. With the addition of a few optical components to the system, the return beams can be recombined and sent to a single detector. In the detector there is a separate frequency component for both the right and left hand component of each ray. These can be separated electronically to provide a reference and probe signal for each ray; the difference in phase between the two signals is twice the Faraday rotation angle.

  10. Magnetic diagnostics for the lithium tokamak experiment.

    PubMed

    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.

  11. RF Wave Propagation and Scattering in Tokamaks

    NASA Astrophysics Data System (ADS)

    Horton, Wendell; Goniche, Marc; Arefiev, Alex; Peysson, Yves; Ekedahl, Annika; InstituteFusion Studies Collaboration; IRFM CEA Collaboration

    2016-10-01

    The propagation, scattering and absorption of the lower hybrid and electron cyclotron RF waves used to control fusion plasmas is reviewed. Drift wave turbulence driven by the steep ion and electron temperature gradients in H-mode divertor tokamaks produces strong scattering of the RF waves used for heating and plasma currents drive Both the 3-5GHz lower-hybrid (LH) and the 170GHZ electron cyclotron (EC) waves experience scattering and diffraction as propagating through the statistically complex density of the plasma. Ray equations are used to calculate the spread of the rays and the associated change in the parallel phase, polarization and group velocity of the RF waves in the propagation through the fusion plasma. A Fokker Planck equation for the phase space of the RF plasmons is one method to describe the spread of the RF wave power in the complex geometry of a divertor tokamak using the ray tracing codes. The evolution of the electron distribution function from the resonant electron-wave interactions is summarized for several scenarios. The resulting X-ray spectrum is broaden giving better agreement with the measured X-ray spectrum than that calculated in the absence of the turbulent scattering of the RF waves. M. Goniche et al., and Tore Supra Team, Phys. Plasmas 21, 2014.

  12. Rotation driven by fast ions in tokamaks

    SciTech Connect

    Thyagaraja, A.; Schwander, F.; McClements, K. G.

    2007-11-15

    Collective fast ion effects on flows in tokamaks are investigated analytically and numerically. A general analysis of noncollisional electrodynamic momentum transfer from fast ions to bulk plasma is presented, with polarization effects and dissipation in the bulk plasma taken into account. The analysis is illustrated using idealized simulations of fast ion orbits and radial electric fields in the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes, R. J. Akers, L. C. Appel et al., Nucl. Fusion 41, 1423 (2001)], the Joint European Torus (JET) [P. H. Rebut et al., Nucl. Fusion 25, 1011 (1985)], and ITER [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Controlled Fusion 44, 519 (2002)]. In the MAST simulation, prompt losses of beam ions injected counter to the plasma current drive up a radial electric field that saturates at a level such that beam ions subsequently injected are confined electrostatically. Although the actual radial electric fields in counterinjected MAST discharges are lower than this, the scenario explored in the simulation would be approached in MAST plasmas with sufficiently low collisionality. The JET simulation, although unrealistic, shows that a similar process could be driven by losses of fusion {alpha}-particles from a burning plasma. Test-particle simulations of {alpha}-particles in ITER suggest that performance-limiting instabilities such as neoclassical tearing modes and resistive wall modes could be affected significantly by flows associated with radial fast particle currents.

  13. Rotation driven by fast ions in tokamaks

    NASA Astrophysics Data System (ADS)

    Thyagaraja, A.; Schwander, F.; McClements, K. G.

    2007-11-01

    Collective fast ion effects on flows in tokamaks are investigated analytically and numerically. A general analysis of noncollisional electrodynamic momentum transfer from fast ions to bulk plasma is presented, with polarization effects and dissipation in the bulk plasma taken into account. The analysis is illustrated using idealized simulations of fast ion orbits and radial electric fields in the Mega-Ampère Spherical Tokamak (MAST) [A. Sykes, R. J. Akers, L. C. Appel et al., Nucl. Fusion 41, 1423 (2001)], the Joint European Torus (JET) [P. H. Rebut et al., Nucl. Fusion 25, 1011 (1985)], and ITER [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Controlled Fusion 44, 519 (2002)]. In the MAST simulation, prompt losses of beam ions injected counter to the plasma current drive up a radial electric field that saturates at a level such that beam ions subsequently injected are confined electrostatically. Although the actual radial electric fields in counterinjected MAST discharges are lower than this, the scenario explored in the simulation would be approached in MAST plasmas with sufficiently low collisionality. The JET simulation, although unrealistic, shows that a similar process could be driven by losses of fusion α-particles from a burning plasma. Test-particle simulations of α-particles in ITER suggest that performance-limiting instabilities such as neoclassical tearing modes and resistive wall modes could be affected significantly by flows associated with radial fast particle currents.

  14. Predicting temperature and density profiles in tokamaks

    SciTech Connect

    Bateman, G.; Kritz, A.H.; Kinsey, J.E.; Redd, A.J.; Weiland, J.

    1998-05-01

    A fixed combination of theory-based transport models, called the Multi-Mode Model, is used in the BALDUR [C. E. Singer {ital et al.}, Comput. Phys. Commun. {bold 49}, 275 (1988)] transport simulation code to predict the temperature and density profiles in tokamaks. The choice of the Multi-Mode Model has been guided by the philosophy of using the best transport theories available for the various modes of turbulence that dominate in different parts of the plasma. The Multi-Mode model has been found to provide a better match to temperature and density profiles than any of the other theory-based models currently available. A description and partial derivation of the Multi-Mode Model is presented, together with three new examples of simulations of the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire {ital et al.}, Phys. Plasmas {bold 2}, 2176 (1995)]. The first simulation shows the strong effect of recycling on the ion temperature profile in TFTR supershot simulations. The second simulation explores the effect of a plasma current ramp{emdash}where the plasma energy content changes slowly on the energy confinement time scale. The third simulation shows that the Multi-Mode Model reproduces the experimentally measured profiles when tritium is used as the hydrogenic isotope in L-mode (low confinement mode) plasmas. {copyright} {ital 1998 American Institute of Physics.}

  15. The Spherical Tokamak MEDUSA for Costa Rica

    NASA Astrophysics Data System (ADS)

    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<0.14m, a<0.10m, BT<0.5T, Ip<40kA, 3ms pulse)[1] is in a process of donation to Costa Rica Institute of Technology. The main objective of MEDUSA is to train students in plasma physics /technical related issues which will help all tasks of the very low aspect ratio stellarator SCR-1(A≡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

  16. Gyrokinetic simulation of microturbulence in EAST tokamak

    NASA Astrophysics Data System (ADS)

    Xiao, Yong; Zhang, Taige; Zhao, Chen

    2014-10-01

    A complete understanding of anomalous transport is critical for designing future magnetic fusion reactors. It is generally accepted that the micro-scale turbulence leads to anomalous transport. For low beta toroidal plasmas, the electrostatic modes may dominate and ion temperature gradient (ITG) mode and trapped electron mode (TEM) are two very important candidates accounting for ion and electron turbulent transport respectively. Recently the massively parallel gyrokinetic simulation has emerged as a major tool to investigate the nonlinear physics of the turbulent transport. The newly-developed capabilities enable the gyrokinetic code GTC to simulate the turbulent transport for real tokamak plasma shape and profiles. These capabilities include a new gyrokinetic Poisson solver and zonal flow solver suitable for general plasma shape and profiles, improvements on the conventional four-point gyroaverage and newly-developed nonuniform initial marker loading. The GTC code is now able to import experimental plasma profiles and equilibrium magnetic field that come from the EFIT or TRANSP equilibrium reconstruction. Linear and nonlinear gyrokinetic simulations are carried out with the new capabilities in GTC for the electron coherent mode (ECM) recently observed in the EAST tokamak (EAST shot # 38300). We found that in the pedestal region with strong electron temperature gradient, the unstable waves propagate in the electron diamagnetic direction, showing a trapped electron mode (TEM) feature. It is also found in the collisionless limit, the linear mode frequency is higher than that from the experiment.

  17. Constrained ripple optimization of Tokamak bundle divertors

    SciTech Connect

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

    1983-02-01

    Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.

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

    SciTech Connect

    Jassby, D.L.

    1986-10-01

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

  19. System studies for quasi-steady-state advanced physics tokamak

    SciTech Connect

    Reid, R.L.; Peng, Y.K.M.

    1983-11-01

    Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated.

  20. Nonlinear stabilization of tokamak microturbulence by fast ions.

    PubMed

    Citrin, J; Jenko, F; Mantica, P; Told, D; Bourdelle, C; Garcia, J; Haverkort, J W; Hogeweij, G M D; Johnson, T; Pueschel, M J

    2013-10-11

    Nonlinear electromagnetic stabilization by suprathermal pressure gradients found in specific regimes is shown to be a key factor in reducing tokamak microturbulence, augmenting significantly the thermal pressure electromagnetic stabilization. Based on nonlinear gyrokinetic simulations investigating a set of ion heat transport experiments on the JET tokamak, described by Mantica et al. [Phys. Rev. Lett. 107, 135004 (2011)], this result explains the experimentally observed ion heat flux and stiffness reduction. These findings are expected to improve the extrapolation of advanced tokamak scenarios to reactor relevant regimes.

  1. Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

    NASA Astrophysics Data System (ADS)

    Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; Holcomb, C. T.; Lao, L. L.; McKee, G. R.; Meneghini, O.; Staebler, G. M.; Grierson, B. A.; Qian, J. P.; Solomon, W. M.; Turnbull, A. D.; Holland, C.; Guo, W. F.; Ding, S. Y.; Pan, C. K.; Xu, G. S.; Wan, B. N.

    2016-06-01

    Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN , despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.

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

    NASA Astrophysics Data System (ADS)

    Nam, Y. U.; Chung, J.

    2010-10-01

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

  3. A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak.

    PubMed

    Ren, J; Zuo, G Z; Hu, J S; Sun, Z; Yang, Q X; Li, J G; Zakharov, L E; Xie, H; Chen, Z X

    2015-02-01

    A program involving the extensive and systematic use of lithium (Li) as a "first," or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak-both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.

  4. MHD analysis of edge instabilities in the JET tokamak

    NASA Astrophysics Data System (ADS)

    Perez von Thun, Christian Pedro

    2004-03-01

    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 research is the study of the equilibrium and stability of a plasma with respect to large scale displacements. In a tokamak, several instabilities can occur. A class of edge instabilities that occur in the high confinement regime, H-mode, have been called Edge Localised Modes (ELMs). ELMs are relaxation oscillations that cause quasiperiodic energy and particle losses out of the confined plasma into the scrape-off layer. These losses are of concern for future burning fusion plasmas, such as ITER, due to the large transient heat loads expected on plasma facing components in contact with the scrape-off layer. These heat loads may reduce the target lifetime below tolerable levels. Although the existence of ELMs has been known for many years, their physics is not well understood yet. Much effort has been spent world-wide in an attempt to improve the understanding of these instabilities. A review of the present state of ELM research is given. Empirically, at least three types of ELMs have been identified, which are normally classified as type-I, type-II and type-III ELMs. From the point of view of plasma stability, research has increasingly focussed on the role of certain MHD instabilities, namely (finite-n) ballooning and kink (peeling) modes, as well as coupled ballooning-kink modes, leading to the proposition of a theoretical model called the peeling-ballooning cycle. This thesis presents new insight into ELMs obtained from the analysis of experimental data in the JET tokamak, and compares the observations with present theoretical ELM models. Low frequency coherent type-I ELM precursor modes have been identified. Their properties are studied in detail. Precursors with low toroidal

  5. An advanced plasma control system for the DIII-D tokamak

    SciTech Connect

    Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J. ); Lazarus, E. )

    1991-11-01

    An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as {beta}{sub p}, {ell}{sub i} and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 {mu}s intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 {mu}s.

  6. QUIESCENT DOUBLE BARRIER H-MODE PLASMAS IN THE DIII-D TOKAMAK

    SciTech Connect

    K.H. BURRELL; M.E. AUSTIN; D.P. BRENNAN; J.C. DeBOO; E.J. DOYLE; C. FENZI; C. FUCHS; P. GOHIL; R.J. GROEBNER; L.L. LAO; T.C. LUCE; M.A. MAKOWSKI; G.R. McKEE; R.A. MOYER; C.C. PETTY; M. PORKOLAB; C.L.RETTIG; T.L. RHODES; J.C. ROST; B.W. STALLARD; E.J. STRAIT; E.J. SYNAKOWSKI; M.R. WADE; J.G. WATKINS; W.P. WEST

    2000-11-01

    High confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J.L. Luxon, et al., Plasma Phys. and Contr. Nucl. Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987) Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation which is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 seconds or >25 energy confinement times {tau}{sub E}), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved {beta}{sub N} H{sub 89} = 7 for up to 5 times the {tau}{sub E} of 150 ms. The {beta}{sub N} H{sub 89} values of 7 substantially exceed the value of 4 routinely achieved in standard ELMing H-mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H-mode is possible because of the presence of an edge magnetic hydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.

  7. ECH by FEL and gyrotron sources on the Microwave Tokamak Experiment (MTX) tokamak

    SciTech Connect

    Stallard, B.W.; Turner, W.C.; Allen, S.L.; Byers, J.A.; Felker, B.; Fenstermacher, M.E.; Ferguson, S.W.; Hooper, E.G.; Thomassen, K.I.; Throop, A.L. ); Makowski, M.A. )

    1990-08-09

    The Microwave Tokamak Experiment (MTX) at LLNL is studying the physics of intense pulse ECH is a high-density tokamak plasma using a microwave FEL. Related technology development includes the FEL, a windowless quasi-optical transmission system, and other microwave components. Initial plasma experiments have been carried out at 140 GHz with single rf pulses generated using the ETA-II accelerator and the ELF wiggler. Peak power levels up to 0.2 GW and pulse durations up to 10 ns were achieved for injection into the plasma using as untapered wiggler. FEL pulses were transmitted over 33 m from the FEL to MTX using six mirrors mounted in a 50-cm-diam evacuated pipe. Measurements of the microwave beam and transmission through the plasma were carried out. For future rapid pulse experiments at high average power (4 GW peak power, 5kHz pulse rate, and {bar P} > 0.5 MW) using the IMP wiggler with tapered magnetic field, a gyrotron (140 GHz, 400 kW cw or up to 1 MW short pulse) is being installed to drive the FEL input or to directly heat the tokamak plasma at full gyrotron power. Quasi-optic techniques will be used to couple the gyrotron power. For direct plasma heating, the gyrotron will couple into the existing mirror transport system. Using both sources of rf generation, experiments are planned to investigate intense pulse absorption and tokamak physics, such as the ECH of a pellet-fueled plasma and plasma control using localized heating. 12 refs., 9 figs.

  8. ECH by FEL and gyrotron sources on the Microwave Tokamak Experiment (MTX) tokamak

    NASA Astrophysics Data System (ADS)

    Stallard, B. W.; Turner, W. C.; Allen, S. L.; Byers, J. A.; Felker, B.; Fenstermacher, M. E.; Ferguson, S. W.; Hooper, E. G.; Thomassen, K. I.; Throop, A. L.

    1990-08-01

    The Microwave Tokamak Experiment (MTX) at LLNL is studying the physics of intense pulse ECH is a high-density tokamak plasma using a microwave FEL. Related technology development includes the FEL, a windowless quasi-optical transmission system, and other microwave components. Initial plasma experiments have been carried out at 140 GHz with single RF pulses generated using the ETA-2 accelerator and the ELF wiggler. Peak power levels up to 0.2 GW and pulse durations up to 10 ns were achieved for injection into the plasma using as untapered wiggler. FEL pulses were transmitted over 33 m from the FEL to MTX using six mirrors mounted in a 50 cm diam evacuated pipe. Measurements of the microwave beam and transmission through the plasma were carried out. For future rapid pulse experiments at high average power (4 GW peak power, 5 kHz pulse rate, and bar P is greater than 0.5 MW) using the IMP wiggler with tapered magnetic field, a gyrotron (140 GHz, 400 kW CW or up to 1 MW short pulse) is being installed to drive the FEL input or to directly heat the tokamak plasma at full gyrotron power. Quasi-optic techniques will be used to couple the gyrotron power. For direct plasma heating, the gyrotron will couple into the existing mirror transport system. Using both sources of RF generation, experiments are planned to investigate intense pulse absorption and tokamak physics, such as the ECH of a pellet-fueled plasma and plasma control using localized heating.

  9. Tokamak reactor cost model based on STARFIRE/WILDCAT costing

    SciTech Connect

    Evans, K. Jr.

    1983-03-01

    A cost model is presented which is useful for survey and comparative studies of tokamak reactors. The model is heavily based on STARFIRE and WILDCAT costing guidelines, philosophies, and procedures and reproduces the costing for these devices quite accurately.

  10. On Stochastic Control of Tokamak and Artificial Intelligence

    NASA Astrophysics Data System (ADS)

    Rastovic, Danilo

    2007-12-01

    Instead of the theory of invariant manifolds and entropy reduction, the theory of fractional Brownian motions and artificiall neural networks is used for description of advanced methods for control of tokamak plasma behaviour.

  11. TFTR/JET INTOR workshop on plasma transport tokamaks

    SciTech Connect

    Singer, C.E.

    1985-01-01

    This report summarizes the proceedings of a Workshop on transport models for prediction and analysis of tokamak plasma confinement. Summaries of papers on theory, predictive modeling, and data analysis are included.

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

    NASA Astrophysics Data System (ADS)

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

    2016-09-01

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

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

    SciTech Connect

    Marinak, M. )

    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. Experimental observations and modelling of intrinsic rotation reversals in tokamaks

    NASA Astrophysics Data System (ADS)

    Camenen, Y.; Angioni, C.; Bortolon, A.; Duval, B. P.; Fable, E.; Hornsby, W. A.; McDermott, R. M.; Na, D. H.; Na, Y.-S.; Peeters, A. G.; Rice, J. E.

    2017-03-01

    The progress made in understanding spontaneous toroidal rotation reversals in tokamaks is reviewed and current ideas to solve this ten-year-old puzzle are explored. The paper includes a summarial synthesis of the experimental observations in AUG, C-Mod, KSTAR, MAST and TCV tokamaks, reasons why turbulent momentum transport is thought to be responsible for the reversals, a review of the theory of turbulent momentum transport and suggestions for future investigations.

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

    SciTech Connect

    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

  16. Global MHD Mode Stabilization and Control for Tokamak Disruption Avoidance

    NASA Astrophysics Data System (ADS)

    Sabbagh, S. A.; Berkery, J. W.; Bialek, J. M.; Hanson, J. M.; Park, Y. S.; Bell, R. E.; Gates, D. A.; Gerhardt, S. P.; Goumiri, I.; Grierson, B.; Holcomb, C.

    2015-11-01

    The near-complete elimination of plasma disruptions in fusion-producing tokamaks is the present ``grand challenge'' for stability research. Meeting this goal requires multiple approaches, important components of which are prediction, stabilization, and control of global MHD instabilities. Research on NSTX and its upgrade is synergizing these elements to make quantified progress on this challenge. Initial results from disruption characterization and prediction analyses describe physical disruption event chains in NSTX. Analysis of NSTX and DIII-D experiments show that stabilization of global modes is dominated by precession drift and bounce orbit resonances respectively. Stability therefore depends on the plasma rotation profile. A model-based rotation profile controller for NSTX-U using both neutral beams and neoclassical toroidal viscosity is shown in simulations to evolve profiles away from unstable states. Active RWM control is addressed using dual field component sensor feedback and a model-based RWM state-space controller. Comparison of measurements and synthetic diagnostics is examined for off-normal event handling. A planned 3D coil system upgrade can allow RWM control close to the ideal n = 1 with-wall limit. Supported by US DOE Contracts DE-FG02-99ER54524 and DE-AC02-09CH11466.

  17. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    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.

  18. Control of asymmetric magnetic perturbations in tokamaks.

    PubMed

    Park, Jong-Kyu; Schaffer, Michael J; Menard, Jonathan E; Boozer, Allen H

    2007-11-09

    The sensitivity of tokamak plasmas to very small deviations from the axisymmetry of the magnetic field |deltaB/B| approximately 10{-4} is well known. What was not understood until very recently is the importance of the perturbation to the plasma equilibrium in assessing the effects of externally produced asymmetries in the magnetic field, even far from a stability limit. DIII-D and NSTX experiments find that when the deleterious effects of asymmetries are mitigated, the external asymmetric field was often made stronger and had an increased interaction with the magnetic field of the unperturbed equilibrium. This Letter explains these counterintuitive results. The explanation using ideal perturbed equilibria has important implications for the control of field errors in all toroidal plasmas.

  19. Transport bifurcation in a rotating tokamak plasma.

    PubMed

    Highcock, E G; Barnes, M; Schekochihin, A A; Parra, F I; Roach, C M; Cowley, S C

    2010-11-19

    The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

  20. Fast tomographic methods for the tokamak ISTTOK

    NASA Astrophysics Data System (ADS)

    Carvalho, P. J.; Thomsen, H.; Gori, S.; Toussaint, U. v.; Weller, A.; Coelho, R.; Neto, A.; Pereira, T.; Silva, C.; Fernandes, H.

    2008-04-01

    The achievement of long duration, alternating current discharges on the tokamak IST-TOK requires a real-time plasma position control system. The plasma position determination based on magnetic probes system has been found to be inadequate during the current inversion due to the reduced plasma current. A tomography diagnostic has been therefore installed to supply the required feedback to the control system. Several tomographic methods are available for soft X-ray or bolo-metric tomography, among which the Cormack and Neural networks methods stand out due to their inherent speed of up to 1000 reconstructions per second, with currently available technology. This paper discusses the application of these algorithms on fusion devices while comparing performance and reliability of the results. It has been found that although the Cormack based inversion proved to be faster, the neural networks reconstruction has fewer artifacts and is more accurate.

  1. Anisotropic pressure tokamak equilibrium and stability considerations

    SciTech Connect

    Salberta, E.R.; Grimm, R.C.; Johnson, J.L.; Manickam, J.; Tang, W.M.

    1987-02-01

    Investigation of the effect of pressure anisotropy on tokamak equilibrium and stability is made with an MHD model. Realistic perpendicular and parallel pressure distributions, P/sub perpendicular/(psi,B) and P/sub parallel/(psi,B), are obtained by solving a one-dimensional Fokker-Planck equation for neutral beam injection to find a distribution function f(E, v/sub parallel//v) at the position of minimum field on each magnetic surface and then using invariance of the magnetic moment to determine its value at each point on the surface. The shift of the surfaces of constant perpendicular and parallel pressure from the flux surfaces depends strongly on the angle of injection. This shift explains the observed increase or decrease in the stability conditions. Estimates of the stabilizing effect of hot trapped ions indicates that a large fraction must be nonresonant and thus decoupled from the bad curvature before it becomes important.

  2. Status of ECRH project on EAST Tokamak

    SciTech Connect

    Wang, Xiaojie; Liu, Fukun; Shan, Jiafang; Xu, Handong; Wu, Dajun; Li, Bo; Zhang, Jiang; Huang, Yiyun; Tang, Yunying; Xu, Weiye; Hu, Huaichuan; Wang, Jian; Xu, Li; Wei, Wei

    2014-02-12

    A 140GHz electron cyclotron resonance heating and current drive (EC H and CD) project for EAST Tokamak is launched in 2011 with a total power of 4MW and pulse length of 100 s. The main objectives of the system are to provide central H and CD, assist start-up and control of MHD activities. The system comprises four gyrotrons each with nominal output power of 1MW at 140GHz. The RF power, transmitted through four evacuated corrugated waveguides will be injected into plasma from the low field side (radial port). The front steering equatorial launcher directs the RF beam over ±25° toroidally and scans over 38° poloidally. At present, the construction of the first 1MW system is undergoing for the expected campaign in the end of 2013. In this paper, the current status of the development and the design of the 140-GHz ECRH system are presented.

  3. A lithium deposition system for tokamak devices*

    NASA Astrophysics Data System (ADS)

    Graziul, Christopher; Majeski, Richard; Kaita, Robert; Hoffman, Daniel; Timberlake, John; Card, David

    2002-11-01

    The production of a lithium deposition system using commercially available components is discussed. This system is intended to provide a fresh lithium wall coating between discharges in a tokamak. For this purpose, a film 100-200 Å thick is sufficient to ensure that the plasma interacts solely with the lithium. A test system consisting of a lithium evaporator and a deposition monitor has been designed and constructed to investigate deposition rates and coverage. A Thermionics 3kW e-gun is used to rapidly evaporate small amounts of solid lithium. An Inficon XTM/2 quartz deposition monitor then measures deposition rate at varying distances, positions and angles relative to the e-gun crucible. Initial results from the test system will be presented. *Supported by US DOE contract #DE-AC02-76CH-03073

  4. Argonne Plasma Engineering Experiment (APEX) Tokamak

    SciTech Connect

    Norem, J.H.; Balka, L.J.; Kulovitz, E.E.; Magill, S.R.; McGhee, D.G.; Moretti, A.; Praeg, W.F.

    1981-03-01

    The Argonne Plasma Engineering Experiment (APEX) Tokamak was designed to provide hot plasmas for reactor-relevant experiments with rf heating (current drive) and plasma wall experiments, principally in-situ low-Z wall coating and maintenance. The device, sized to produce energetic plasmas at minimum cost, is small (R = 51 cm, r = 15 cm) but capable of high currents (100 kA) and long pulse durations (100 ms). A design using an iron central core with no return legs, pure tension tapewound toroidal field coils, digital radial position control, and UHV vacuum technology was used. Diagnostics include monochrometers, x-ray detectors, and a microwave interferometer and radiometer for density and temperature measurements. Stable 100 ms shots were produced with electron temperatures in the range 500 to 1000 eV. Initial results included studies of thermal desorption and recoating of wall materials.

  5. Nusselt number scaling in tokamak plasma turbulence

    SciTech Connect

    Takeda, K.; Benkadda, S.; Hamaguchi, S.; Wakatani, M.

    2005-05-15

    Anomalous heat transport caused by ion temperature gradient (ITG) driven turbulence in tokamak plasmas is evaluated from numerical simulations of the two-dimensional (2D) partial-differential equations of the ITG model and of a reduced 1D version derived from a quasilinear approximation. In the strongly turbulent state, intermittent bursts of thermal transport are observed in both cases. In the strongly turbulent regime, the reduced model as well as the direct numerical simulation show that the Nusselt number Nu (normalized heat flux) scales with the normalized ion pressure gradient K{sub i} as Nu{proportional_to}K{sub i}{sup 1/3}. Since the Rayleigh number for ITG turbulence is proportional to K{sub i}, the Nusselt number scaling for ITG turbulence is thus similar to the classical thermal transport scaling for Rayleigh-Benard convections in neutral fluids.

  6. Cooldown of the Compact Ignition Tokamak

    SciTech Connect

    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.

  7. Tearing mode analysis in tokamaks, revisited

    SciTech Connect

    Nishimura, Y.; Callen, J.D.; Hegna, C.C.

    1998-12-01

    A new {Delta}{sup {prime}} shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio ({epsilon}{le}0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of the finite pressure term. Numerical results compare favorably with Furth {ital et al.} [H. P. Furth {ital et al.}, Phys. Fluids {bold 16}, 1054 (1973)] results. The effects of finite pressure, which are shown to decrease {Delta}{sup {prime}}, are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric elements, stabilizes the tearing mode significantly, even in a low-{beta} regime before the toroidal magnetic curvature effects come into play. {copyright} {ital 1998 American Institute of Physics.}

  8. Equilibrium properties on the EAST superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Qian, J. P.; Wan, B. N.; Lao, L. L.; Shen, B.; Sabbagh, S. A.; Menard, J.; Sun, Y. W.; Duan, Y. M.; Li, J. H.; Xiao, B. J.; Gong, X. Z.; Gong

    2009-06-01

    The Experimental Advanced Superconducting Tokamak (EAST) has a major radius of R0 = 1.75 m and a midplane halfwidth of 0.5 m. It has been operated with a toroidal magnetic field B0 = 2 T and Ip ≤ 500 kA. The evolution of the plasma equilibrium is analysed between discharges by Equilibrium Fitting Code (EFIT). Limiter, single-null and double-null diverted configurations have been produced. A plasma elongation in the range 1.3 ≤ κ ≤ 1.9 and a triangularity in the range 0.1 ≤ δ ≤ 0.55 have been sustained. The operation space of elongated discharges is also presented based on the EAST database.

  9. Neutral Beam Injection in the Electric Tokamak

    NASA Astrophysics Data System (ADS)

    Gourdain, P.-A.; Carter, T. A.; Gauvreau, J.-L.; Grossman, A.; Lafonteese, D. J.; Pace, D. C.; Schmitz, L. W.; Taylor, R. J.; White, A. E.; Yates, T. F.

    2004-11-01

    The Electric Tokamak (ET) at UCLA (Bt=0.25T, R=5m, a=1m, Te(0)=300eV, tau(0)=1s) is now running long shots (5s). A new development program was started last year to include a neutral beam in the daily operations of the machine. As a result, a 10kV neutral beam injector was built to deal with plasma and measurement issues. The design and parameters of the beam are discussed. The source is based on an RF generated plasma, with a single extraction grid providing an accel-decel configuration. Plasma neutralization efficiency is also presented. Co- or counter injection is now possible using a single beam. The construction of a second beam is planned for simultaneous co- and counter injections for toroidal momentum input control. Plasma toroidal and poloidal rotation, particle diffusion and current drive effects will be presented.

  10. Fast tomographic methods for the tokamak ISTTOK

    SciTech Connect

    Carvalho, P. J.; Coelho, R.; Neto, A.; Pereira, T.; Silva, C.; Fernandes, H.; Gori, S.; Toussaint, U. v.

    2008-04-07

    The achievement of long duration, alternating current discharges on the tokamak IST-TOK requires a real-time plasma position control system. The plasma position determination based on magnetic probes system has been found to be inadequate during the current inversion due to the reduced plasma current. A tomography diagnostic has been therefore installed to supply the required feedback to the control system. Several tomographic methods are available for soft X-ray or bolo-metric tomography, among which the Cormack and Neural networks methods stand out due to their inherent speed of up to 1000 reconstructions per second, with currently available technology. This paper discusses the application of these algorithms on fusion devices while comparing performance and reliability of the results. It has been found that although the Cormack based inversion proved to be faster, the neural networks reconstruction has fewer artifacts and is more accurate.

  11. Transport Bifurcation in a Rotating Tokamak Plasma

    SciTech Connect

    Highcock, E. G.; Barnes, M.; Schekochihin, A. A.; Parra, F. I.; Roach, C. M.; Cowley, S. C.

    2010-11-19

    The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

  12. Diagnostics modules for tokamak disruption experiments

    SciTech Connect

    Nahm, M.L.; Buchanan, C.D.; Bourham, M.A.; Gilligan, J.G.

    1994-11-01

    Diagnostic modules equipped with various sensors can provide useful information on key parameters for disruption events, e.g. energy deposition, vapor shielding effect, plasma pressure and force distribution. The modules are, basically, DIMES samples (Divertor Materials Evaluation System) equipped with sensors, coupled to digitizing units and interfaced to a data acquisition system. The DIMES samples are part of the lower diverter diagnostics on the DIII-D tokamak. Three top-cap prototype diagnostics modules have been designed and fabricated. The initial testing and calibration have been performed using the SIRENS plasma gun at an energy deposition of 1 to 12 MJ/m{sup 2} over 0.1 to 1.0 ms, with a plasma pressure >100 MPa.

  13. Passive runaway electron suppression in tokamak disruptions

    SciTech Connect

    Smith, H. M.; Helander, P.

    2013-07-15

    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.

  14. TRAIL: a tokamak rail gun limiter

    SciTech Connect

    Yu, W.S.; Powell, J.R.; Usher, J.L.

    1980-01-01

    An attractive new limiter concept is investigated. The TRAIL (Tokamak Rail Gun Limiter) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled, after cooling, to the injector in an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state-of-the-art (several km/sec). Accelerators injecting pellets at approx. 1 km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  15. Vertically stabilized elongated cross-section tokamak

    DOEpatents

    Sheffield, George V.

    1977-01-01

    This invention provides a vertically stabilized, non-circular (minor) cross-section, toroidal plasma column characterized by an external separatrix. To this end, a specific poloidal coil means is added outside a toroidal plasma column containing an endless plasma current in a tokamak to produce a rectangular cross-section plasma column along the equilibrium axis of the plasma column. By elongating the spacing between the poloidal coil means the plasma cross-section is vertically elongated, while maintaining vertical stability, efficiently to increase the poloidal flux in linear proportion to the plasma cross-section height to achieve a much greater plasma volume than could be achieved with the heretofore known round cross-section plasma columns. Also, vertical stability is enhanced over an elliptical cross-section plasma column, and poloidal magnetic divertors are achieved.

  16. Nonlinear lower hybrid modeling in tokamak plasmas

    SciTech Connect

    Napoli, F.; Schettini, G.; Castaldo, C.; Cesario, R.

    2014-02-12

    We present here new results concerning the nonlinear mechanism underlying the observed spectral broadening produced by parametric instabilities occurring at the edge of tokamak plasmas in present day LHCD (lower hybrid current drive) experiments. Low frequency (LF) ion-sound evanescent modes (quasi-modes) are the main parametric decay channel which drives a nonlinear mode coupling of lower hybrid (LH) waves. The spectrum of the LF fluctuations is calculated here considering the beating of the launched LH wave at the radiofrequency (RF) operating line frequency (pump wave) with the noisy background of the RF power generator. This spectrum is calculated in the frame of the kinetic theory, following a perturbative approach. Numerical solutions of the nonlinear LH wave equation show the evolution of the nonlinear mode coupling in condition of a finite depletion of the pump power. The role of the presence of heavy ions in a Deuterium plasma in mitigating the nonlinear effects is analyzed.

  17. Self-organized stationary states of tokamaks

    NASA Astrophysics Data System (ADS)

    Jardin, Stephen

    2015-11-01

    We report here on a nonlinear mechanism that forms and maintains a self-organized stationary (sawtooth free) state in tokamaks. This process was discovered by way of extensive long-time simulations using the M3D-C1 3D extended MHD code in which new physics diagnostics have been added. It is well known that most high-performance modes of tokamak operation undergo ``sawtooth'' cycles, in which the peaking of the toroidal current density triggers a periodic core instability which redistributes the current density. However, certain modes of operation are known, such as the ``hybrid'' mode in DIII-D, ASDEX-U, JT-60U and JET, and the long-lived modes in NSTX and MAST, which do not experience this cycle of instability. Empirically, it is observed that these modes maintain a non-axisymmetric equilibrium which somehow limits the peaking of the toroidal current density. The physical mechanism responsible for this has not previously been understood, but is often referred to as ``flux-pumping,'' in which poloidal flux is redistributed in order to maintain q0 >1. In this talk, we show that in long-time simulations of inductively driven plasmas, a steady-state magnetic equilibrium may be obtained in which the condition q0 >1 is maintained by a dynamo driven by a stationary marginal core interchange mode. This interchange mode, unstable because of the pressure gradient in the ultra-low shear region in the center region, causes a (1,1) perturbation in both the electrostatic potential and the magnetic field, which nonlinearly cause a (0,0) component in the loop voltage that acts to sustain the configuration. This hybrid mode may be a preferred mode of operation for ITER. We present parameter scans that indicate when this sawtooth-free operation can be expected.

  18. The ARIES Advanced and Conservative Tokamak Power Plant Study

    DOE PAGES

    Kessel, C. E; Tillak, M. S; Najmabadi, F.; ...

    2015-12-22

    Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦtotal N of 5.75, an H98 of 1.65, anmore » n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦtotalN of 2.5, an H₉₈ of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less

  19. The ARIES Advanced and Conservative Tokamak Power Plant Study

    SciTech Connect

    Kessel, C. E; Tillak, M. S; Najmabadi, F.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; EL-Guebaly, L.; Blanchard, J. P.; Martin, C. J.; Mynsburge, L.; Humrickhouse, P.; Rensink, M. E.; Rognlien, T. D.; Yoda, M.; Abdel-Khalik, S. I.; Hageman, M. D.; Mills, B. H.; Rader, J. D.; Sadowski, D. L.; Snyder, P. B.; St. John, H.; Turnbull, A. D.; Waganer, L. M.; Malang, S.; Rowcliffe, A. F.

    2015-12-22

    Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦtotal N of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦtotalN of 2.5, an H₉₈ of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.

  20. The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study

    SciTech Connect

    Kessel, C. E.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Tillack, M. S.; Najmabadi, F.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; El-Guebaly, L.; Blanchard, J. P.; Martin, C. J.; Mynsburge, L.; Humrickhouse, P.; Rensink, M. E.; Rognlien, T. D.; Yoda, M.; Abdel-Khalik, S. I.; Hageman, M. D.; Mills, B. H.; Radar, J. D.; Sadowski, D. L.; Snyder, P. B.; St. John, H.; Turnbull, A. D.; Waganer, L. M.; Malang, S.; Rowcliffe, A. F.

    2014-03-05

    Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a {beta}N{sup total} of 5.75, H{sub 98} of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m{sup 2}. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a {beta}N{sup total} of 2.5, H{sub 98} of 1.25, n/n{sub Gr} of 1.3, and peak divertor heat flux of 10 MW/m{sup 2}. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m{sup 2}. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.

  1. Lower hybrid system design for the Tokamak physics experiment

    SciTech Connect

    Goranson, P.L.; Conner, D.L.; Swain, D.W.; Yugo, J.J.; Bernabei, S.; Greenough, N.

    1995-12-31

    The lower hybrid (LH) launcher configuration has been redesigned to integrate the functions of the vertical four-way power splitter and the front waveguide array (front array). This permits 256 waveguide channels to be fed by only 64 waveguides at the vacuum window interface. The resulting configuration is a more compact coupler, which incorporates the simplicity of a multijunction coupler while preserving the spectral flexibility of a conventional lower hybrid launcher. Other spin-offs of the redesign are reduction in thermal incompatibility between the front array and vacuum windows, improved maintainability, in situ vacuum window replacement, a reduced number of radio frequency (rf) connections, and a weight reduction of 7300 kg. There should be a significant cost reduction as well. Issues associated with the launcher design and fabrication have been addressed by a research and development program that includes brazing of the front array and testing of the power splitter configuration to confirm that phase errors due to reflections in the shorted splitter legs will not significantly impact the rf spectrum. The Conceptual Design Review requires that radiation levels at the torus radial port mounting flange and outer surface of the toroidal field coils should be sufficiently low to permit hands-on maintenance. Low activation materials and neutron shielding are incorporated in the launcher design to meet these requirements. The launcher is configured to couple 3 MW of steady state LH heating/LH current drive power at 3.7 GHz to the Tokamak Physics Experiment plasma.

  2. SODA: The reduced database for the TdeV tokamak

    SciTech Connect

    Cote, A.; Michaud, D.; Caumartin, J.; de Villers, P.; Gauthier, Y.; Gauvreau, J.; Larsen, J.

    1997-01-01

    SODA which stands for Syst{grave e}me d{close_quote}Organisation des Donn{acute e}es et d{close_quote}Analyse, is a general database for TdeV. SODA has the following goals: to produce a database of a reduced set of physical data; to ensure that these data are validated; to record all the parameters relevant to tokamak operation and experiments; to facilitate the retrieval of data using given selection criteria; and to improve data accessibility and analysis. The relational database ORACLE{trademark} has been chosen to provide flexibility and to accommodate the increasing expectations of the TdeV researchers. In-house expertise allows custom-made tables and centralized data management. In the process of creating SODA several new interfaces for the scientific coordinator, machine operator, and diagnosticians have been added to provide a better definition of the experiment for the archiving system. The database includes the more relevant machine and diagnostic parameters, plasma perturbations (rf, biasing, gas{hor_ellipsis}), mean and standard deviation of physical signals, plasma profiles, and code results (equilibrium{hor_ellipsis}) for selected time windows in a discharge. Users of the X-window interface of SODA are not required to know the database structure or the SQL language. SODA has been operating successfully for over a year and its capabilities are continuously expanding. {copyright} {ital 1997 American Institute of Physics.}

  3. Diagnostic of fusion neutrons on JET tokamak using diamond detector

    SciTech Connect

    Nemtsev, G.; Amosov, V.; Marchenko, N.; Meshchaninov, S.; Rodionov, R.; Popovichev, S.; Collaboration: JET EFDA Conbributors

    2014-08-21

    In 2011-2012, an experimental campaign with a significant yield of fusion neutrons was carried out on the JET tokamak. During this campaign the facility was equipped with two diamond detectors based on natural and artificial CVD diamond. These detectors were designed and manufactured in State Research Center of Russian Federation TRINITI. The detectors measure the flux of fast neutrons with energies above 0.2 MeV. They have been installed in the torus hall and the distance from the center of plasma was about 3 m. For some of the JET pulses in this experiment, the neutron flux density corresponded to the operational conditions in collimator channels of ITER Vertical Neutron Camera. The main objective of diamond monitors was the measurement of total fast neutron flux at the detector location and the estimation of the JET total neutron yield. The detectors operate as threshold counters. Additionally a spectrometric measurement channel has been configured that allowed us to distinguish various energy components of the neutron spectrum. In this paper we describe the neutron signal measuring and calibration procedure of the diamond detector. Fluxes of DD and DT neutrons at the detector location were measured. It is shown that the signals of total neutron yield measured by the diamond detector correlate with signals measured by the main JET neutron diagnostic based on fission chambers with high accuracy. This experiment can be considered as a successful test of diamond detectors in ITER-like conditions.

  4. Polarimetric spectra analysis for tokamak pitch angle measurements

    NASA Astrophysics Data System (ADS)

    Ko, J.; Chung, J.; Lange, A. G. G.; de Bock, M. F. M.

    2013-10-01

    Measurements of the internal magnetic field structures using conventional polarimetric approaches are considered extremely challenging in fusion-reactor environments whereas the information on current density profiles is essential to establish steady-state and advance operation scenarios in such reactor-relevant devices. Therefore, on ITER a hybrid system is proposed for the current density measurements that uses both polarimetry and spectral measurements. The spectrum-based approaches have been tested in the Korea Superconducting Tokamak Advanced Research (KSTAR) during the past two plasma campaigns. As such, KSTAR is a test-bed for the proposed ITER hybrid system. Measurements in the plasma core are based on the motional Stark effect (MSE) spectrum of the neutral beam emission. For the edge profiles, the Zeeman effect (ZE) acting on the lithium emission spectrum of the newly installed (2013) Lithium-beam-diagnostic is exploited. The neutral beam emission spectra, complicated by the multi-ion-source beam injection, are successfully fitted making use of the data provided by the Atomic Data and Analysis Structure (ADAS) database package. This way pitch angle profiles could be retrieved from the beam emission spectra. With the same spectrometer/CCD hardware as on MSE, but with a different wavelength range and different lines of sight, the first ZE spectrum measurements have been made. The Zeeman splitting comparable to and greater than the instrumental broadening has been routinely detected at high toroidal field operations ( ~ 3 Tesla).

  5. Fast pulsed radar reflectometry for the Textor Tokamak

    NASA Astrophysics Data System (ADS)

    Hugenholtz, C. A. J.; Donné, A. J. H.; Elzendoorn, B. S. Q.; van Gorkom, J. C.; Kooijman, W.; van der Laan, H. A.; van de Pol, M. J.; Putter, A. J.; van Ramele, H. J. F.; Smit, D.; de Vries, P. C.; Wijnoltz, F.; Pysik, W.; Waidmann, G.; Ermak, G. P.

    1999-01-01

    Progress in fusion research shows an increasing demand for diagnostics with high temporal and spatial resolution in order to study small and fast phenomena in fusion plasmas. Therefore, a new ten-channel pulsed radar reflectometer is under development for the TEXTOR Tokamak, to measure electron density profiles in the range of 0.4-4×1019m-3 at a high repetition rate of 2 MHz. An additional feature of this diagnostic is the possibility to perform correlation measurements at 10 MHz repetition rate to investigate density fluctuations. The reflectometer will scan the plasma with 1-ns-long radar pulses in the 18-57 GHz range. In spite of the long flight time of the pulses of about 100 ns, the above mentioned high repetition rates could be achieved by time multiplexing the generation and reception of the radar pulses. Temporal selection of the received pulses is performed by fast switching of the local oscillators inputs of the heterodyne receiver, to generate time windows in which the reception of the reflected pulses is expected. An embedded VME controller will manage the system and store the data with a speed of 20 Msamples/s, up to a maximum of 64 Mbyte data per plasma shot. In order to facilitate the handling of such a huge amount of data, an advanced data reduction scheme is being developed. Remote operation with a fast data link from FZ Jülich (Germany) to FOM Nieuwegein (The Netherlands) will be possible.

  6. Remote maintenance concepts for the Compact Ignition Tokamak

    SciTech Connect

    Davis, F.C.; Hager, E.R.

    1988-01-01

    Because deuterium-tritium fuel will be used in the Compact Ignition Tokamak (CIT), remote handling technology is needed to carry out some maintenance operations on the machine. In keeping with the compact, low-cost nature of CIT, remote maintenance is provided only for systems with the highest probability of failure. Remote operations include removing, repairing (if feasible), and replacing such components as thermal protection tiles on the first wall, radio-frequency (rf) heating modules, and diagnostic modules. For maintenance inside the vacuum vessel, major pieces of equipment under development include an articulated boom manipulator with servomanipulators, an inspection manipulator, and special tooling. For maintenance outside the cryostat, remote equipment includes a bridge-mounted manipulator system, equipment for decontamination and hot cell activities, and for handling and packaging solid radioactive waste. The conceptual design phase of the CIT project is nearing completion; research and development activities in support of the project include demonstrations of remote maintenance operations on full-size partial mock-ups. 9 figs.

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

    SciTech Connect

    BURRELL,HK

    2002-11-01

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

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

    SciTech Connect

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

    2000-12-31

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

  9. Supersonic Molecular Beam Injection Effects on Tokamak Plasma Applied Non-axisymmetric Magnetic Perturbation

    NASA Astrophysics Data System (ADS)

    Han, Hyunsun; in, Y.; Jeon, Y. M.; Hahn, S. H.; Lee, K. D.; Nam, Y. U.; Yoon, S. W.

    2016-10-01

    In KSTAR experiments, the change of tokamak plasma behavior by supersonic molecular beam injection (SMBI) was investigated by applying resonant magnetic perturbations(RMP) that could suppress edge localized modes (ELMs). When the SMBI is applied, the symptom representing ELM suppression by RMP is disappeared. The SMBI acts as a cold pulse on the plasma keeping the total confinement engergy constant. However, it makes plasma density increase and change the plasama collisionality which can play a role in the edge-pedestal build-up processing. This work was supported by Project PG1201-2 and the KSTAR research project funded by Korea Ministry of Science, ICT and Future Planning.

  10. Magnetic Flux Reconstruction Methods for Shaped Tokamaks

    NASA Astrophysics Data System (ADS)

    Tsui, Chi-Wa.

    The use of a variational method permits the Grad -Shafranov (GS) equation to be solved by reducing the problem of solving the 2D non-linear partial differential equation to the problem of minimizing a function of several variables. This high speed algorithm approximately solves the GS equation given a pararmeterization of the plasma boundary and the current profile (p^' and FF^' functions). We treat the current profile parameters as unknowns. The goal is to reconstruct the internal magnetic flux surfaces of a tokamak plasma and the toroidal current density profile from the external magnetic measurements. This is a classic problem of inverse equilibrium determination. The current profile parameters can be evaluated by several different matching procedures. We found that the matching of magnetic flux and field at the probe locations using the Biot-Savart law and magnetic Green's function provides a robust method of magnetic reconstruction. The matching of poloidal magnetic field on the plasma surface provides a unique method of identifying the plasma current profile. However, the power of this method is greatly compromised by the experimental errors of the magnetic signals. The Casing Principle (60) provides a very fast way to evaluate the plasma contribution to the magnetic signals. It has the potential of being a fast matching method. We found that the performance of this method is hindered by the accuracy of the poloidal magnetic field computed from the equilibrium solver. A flux reconstruction package have been implemented which integrates a vacuum field solver using a filament model for the plasma, a multi-layer perceptron neural network as a interface, and the volume integration of plasma current density using Green's functions as a matching method for the current profile parameters. The flux reconstruction package is applied to compare with the ASEQ and EFIT data. The results are promising. Also, we found that some plasmas in the tokamak Alcator C-Mod lie

  11. Magnetic control of magnetohydrodynamic instabilities in tokamaks

    NASA Astrophysics Data System (ADS)

    Strait, E. J.

    2015-02-01

    Externally applied, non-axisymmetric magnetic fields form the basis of several relatively simple and direct methods to control magnetohydrodynamic (MHD) instabilities in a tokamak, and most present and planned tokamaks now include a set of non-axisymmetric control coils for application of fields with low toroidal mode numbers. Non-axisymmetric applied fields are routinely used to compensate small asymmetries ( δB /B ˜10-3 to 10-4 ) of the nominally axisymmetric field, which otherwise can lead to instabilities through braking of plasma rotation and through direct stimulus of tearing modes or kink modes. This compensation may be feedback-controlled, based on the magnetic response of the plasma to the external fields. Non-axisymmetric fields are used for direct magnetic stabilization of the resistive wall mode—a kink instability with a growth rate slow enough that feedback control is practical. Saturated magnetic islands are also manipulated directly with non-axisymmetric fields, in order to unlock them from the wall and spin them to aid stabilization, or position them for suppression by localized current drive. Several recent scientific advances form the foundation of these developments in the control of instabilities. Most fundamental is the understanding that stable kink modes play a crucial role in the coupling of non-axisymmetric fields to the plasma, determining which field configurations couple most strongly, how the coupling depends on plasma conditions, and whether external asymmetries are amplified by the plasma. A major advance for the physics of high-beta plasmas ( β = plasma pressure/magnetic field pressure) has been the understanding that drift-kinetic resonances can stabilize the resistive wall mode at pressures well above the ideal-MHD stability limit, but also that such discharges can be very sensitive to external asymmetries. The common physics of stable kink modes has brought significant unification to the topics of static error fields at low

  12. Magnetic control of magnetohydrodynamic instabilities in tokamaks

    SciTech Connect

    Strait, E. J.

    2015-02-15

    Externally applied, non-axisymmetric magnetic fields form the basis of several relatively simple and direct methods to control magnetohydrodynamic (MHD) instabilities in a tokamak, and most present and planned tokamaks now include a set of non-axisymmetric control coils for application of fields with low toroidal mode numbers. Non-axisymmetric applied fields are routinely used to compensate small asymmetries (δB/B∼10{sup −3} to 10{sup −4}) of the nominally axisymmetric field, which otherwise can lead to instabilities through braking of plasma rotation and through direct stimulus of tearing modes or kink modes. This compensation may be feedback-controlled, based on the magnetic response of the plasma to the external fields. Non-axisymmetric fields are used for direct magnetic stabilization of the resistive wall mode—a kink instability with a growth rate slow enough that feedback control is practical. Saturated magnetic islands are also manipulated directly with non-axisymmetric fields, in order to unlock them from the wall and spin them to aid stabilization, or position them for suppression by localized current drive. Several recent scientific advances form the foundation of these developments in the control of instabilities. Most fundamental is the understanding that stable kink modes play a crucial role in the coupling of non-axisymmetric fields to the plasma, determining which field configurations couple most strongly, how the coupling depends on plasma conditions, and whether external asymmetries are amplified by the plasma. A major advance for the physics of high-beta plasmas (β = plasma pressure/magnetic field pressure) has been the understanding that drift-kinetic resonances can stabilize the resistive wall mode at pressures well above the ideal-MHD stability limit, but also that such discharges can be very sensitive to external asymmetries. The common physics of stable kink modes has brought significant unification to the topics of static error

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

    SciTech Connect

    1995-09-22

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

  14. Initial boronization of the DIII-D tokamak

    SciTech Connect

    Phillips, J.; Hodapp, T.; Holtrop, K.; Jackson, G.; Winter, J.

    1991-10-01

    A system has been installed to deposit a thin film of elemental boron to the walls of the D3-D tokamak, in order to reduce the influx of impurities during plasma discharges. Subsequently new regimes of substantially tokamak energy confinement were obtained. The deposition of the boron layer is achieved during a glow-discharge session using a helium-diborane age mixture and a film of {approx}100 nm is deposited. The boronization system includes special storage and handling precautions for the diborane, a delivery and metering system for the glow-discharge, modifications to the tokamak's residual gas analyzer system, and a dedicated system for handling and neutralizing the exhaust gas from the tokamak. Tokamak discharges with similar parameters before and after boronization are used to characterize the effects of the boron film. Nickel has been reduced by a factor of 30, while impurities such as oxygen and carbon are reduced fivefold. A system of pulsing the flow discharge has been developed in order to improve the uniformity of the film applied.

  15. Magnetic Start-up Inducement in NOVILLO Tokamak

    NASA Astrophysics Data System (ADS)

    Melendez-Lugo, Leandro; Chavez, Esteban; Barocio, Samuel R.; Flores, Angel; Cruz, Guillermo J.; Guadalupe Olayo, M.

    2000-10-01

    Most tokamak devices require one or more breakdown assistance systems, such as particles or ECRH radiation, in order to develop plasma discharges by increasing the probabilities of ionization. In smaller tokamaks filament preionization traditionally primes the start-up, reducing the required amounts of loop voltage and OHT flux consumption. Yet, the latter not only may result obtrusive to the operational access but also its incandescence contributes to raise the plasma impurity levels. A new vertical field profile, superimposed to the stray field component parallel to the equilibrium field, has been created in NOVILLO tokamak by means of two additional compensation poloidal coils. Such arrangement of the vertical field profiles, with an 1/R-dependence, has been found adequate to achieve the plasma breakdown without any preionization system. Furthermore, evidence has been gathered indicating that tokamak discharges achieved by vertical stray field modification instead of preionization allow to operate with higher prefill pressure and therefore to create denser plasmas at standard loop voltage levels. A technique like the one presented here may prove suitable to accomplish dependable discharges on large tokamaks while saving costs considerably. Likewise this technique is feasible to be applied successfully in general toroidal gas breakdown and arcing processes.

  16. Tokamak L/H mode transition

    SciTech Connect

    Tsui, K. H.; Navia, C. E.

    2012-01-15

    Through the non field-aligned rotational tokamak equilibrium of a divergence-free plasma flow with a pair of transformed plasma variables w-vector{sub *}=({mu}{rho}){sup 1/2}{nu}-vector and {mu}p{sub *}=({mu}p+w{sub *}{sup 2}/2)[K. H. Tsui, Phys. Plasmas 18, 072502 (2011)], a preliminary understanding of the L/H equilibrium transition is proposed through a feedback cycle, where the higher plasma flux due to external drives enters the rotational Grad-Shafranov equation through the velocity dependent poloidal plasma {beta} to generate the H equilibrium. This H rotational mode has the characteristics of higher normal electric field and plasma pressure. Coupled to the transport properties of E-vector x B-vector drift transport barrier leading to a higher plasma pressure, this makes the H mode a self-sustained equilibrium. The higher plasma {beta} then feeds back to the equilibrium and completes the feedback loop.

  17. Physics aspects of the Compact Ignition Tokamak

    SciTech Connect

    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.

  18. Forced magnetic reconnection in Tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Cole, Andrew Joseph

    This dissertation addresses two related problems in the study of forced magnetic reconnection in Tokamak plasmas. First, a recent controversy concerning a model forced magnetic reconnection problem, the Taylor problem, has been resolved. The criticisms of Ishizawa and Tokuda [21] concerning the original analysis of Hahm and Kulsrud [17] are shown to be unwarranted, both analytically and numerically. Second, one possible reason for the discrepancy between recent experimental [29] and previous theoretical [13] scaling of the critical error-field penetration threshold with device parameters is addressed. The theory in question is entirely based on a single-fluid MHD (magnetohydrodynamical) treatment of the plasma. As is well-known, high temperature plasmas are far better modeled using the drift-MHD ordering.[18] Hence we develop a drift-MHD theory of error-field penetration. Although two new drift-MHD plasma response regimes are identified, the overall threshold scaling with device parameters is not altogether different from that predicted by single-fluid MHD.

  19. Zonal flows in tokamaks with anisotropic pressure

    SciTech Connect

    Ren, Haijun

    2014-04-15

    Zonal flows (ZFs) in a tokamak plasma with anisotropic pressure are investigated. The dynamics of perpendicular and parallel pressures are determined by the Chew-Goldberger-Low double equations and low-β condition is adopted, where β is the ratio of plasma pressure to the magnetic field pressure. The dispersion relation is analytically derived and illustrates two branches of ZFs. The low frequency zonal flow (LFZF) branch becomes unstable when χ, the ratio of the perpendicular pressure to the parallel one, is greater than a threshold value χ{sub c}, which is about 3.8. In the stable region, its frequency increases first and then decreases with increasing χ. For χ = 1, the frequency of LFZF agrees well with the experimental observation. For the instability, the growth rate of LFZF increases with χ. The geodesic acoustic mode branch is shown to be always stable with a frequency increasing with χ. The safety factor is shown to diminish the frequencies of both branches or the growth rate of LFZF.

  20. Tokamak power system studies at ANL

    SciTech Connect

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

    1986-06-01

    The following features, in particular, have been examined: (a) large aspect ratio (A approx. = 6), which may ease maintenance; (b) high beta (..beta.. greater than or equal to 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I approx. = 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors.

  1. Sensitivity of predictive tokamak plasma transport simulations

    SciTech Connect

    Redd, A.J.; Kritz, A.H.; Bateman, G.; Kinsey, J.E.

    1997-06-01

    The sensitivity of our time-dependent simulations of low confinement (L-mode) discharges to variations in initial profiles and time-dependent boundary conditions has been explored. These time-dependent tokamak plasma simulations were performed using a theory-based Multi-mode transport model that includes ion temperature gradient (ITG) and trapped electron modes (TEM), kinetic and resistive ballooning modes and neoclassical modes. The density and temperature profiles predicted in our simulations of L-mode discharges are found to be robust, even with significant variations in the initial or boundary conditions. Although transport associated with a single mode can be strongly affected by local changes in plasma parameters resulting from changes in the boundary conditions, the total transport remains largely unchanged because of compensation by other transport modes. The sensitivity of the predicted temperature and density profiles to a variation in the Multi-mode model is also examined. When the Dominguez-Waltz theory of transport driven by ITG and TEM modes is replaced in the Multi-mode model by the Weiland description, we find that the predictions of the Weiland model more closely match the experimental data. {copyright} {ital 1997 American Institute of Physics.}

  2. Quasicoherent modes on the COMPASS tokamak

    NASA Astrophysics Data System (ADS)

    Melnikov, A. V.; Markovic, T.; Eliseev, L. G.; Adámek, J.; Aftanas, M.; Bilkova, P.; Boehm, P.; Gryaznevich, M.; Imrisek, M.; Lysenko, S. E.; Medvedev, S. Y.; Panek, R.; Peterka, M.; Seidl, J.; Stefanikova, E.; Stockel, J.; Weinzettl, V.; the COMPASS Team

    2015-06-01

    Multiple quasicoherent electromagnetic modes with steady-state frequency and different nature and location were observed in the COMPASS tokamak (R = 0.56 m, = 0.2 m) at Bt = 1.14 T with Co-NBI (PNBI = 0.2-0.5 MW, Eb = 32 keV) at frequencies 5 kHz < f < 250 kHz. Modes were observed in both low and high confinement (L- and H-modes) plasmas. Lower frequency modes with f < 50 kHz were identified as low m tearing and kink MHD modes, while higher frequency modes with 50 kHz < f < 250 kHz were considered as having Alfvénic nature. Unexpectedly, such modes were only observed in the H-mode, both in neutral beam injector-assisted and Ohmic, so the mode driving force is not yet clear. Using the linear MHD code KINX, we initially identified the observed mode with a ballooning structure is as beta induced Alfvén eigenmode (BAE) with m, n < 5, while an antiballooning mode is initially identified as toroidal Alfvén eigenmode (TAE) with m, n < 9.

  3. Thermo-Oxidation of Tokamak Carbon Dust

    SciTech Connect

    J.W. Davis; B.W.N. Fitzpatrick; J.P. Sharpe; A.A. Haasz

    2008-04-01

    The oxidation of dust and flakes collected from the DIII-D tokamak, and various commercial dust specimens, has been measured at 350 ºC and 2.0 kPa O2 pressure. Following an initial small mass loss, most of the commercial dust specimens showed very little effect due to O2 exposure. Similarly, dust collected from underneath DIII-D tiles, which is thought to comprise largely Grafoil™ particulates, also showed little susceptibility to oxidation at this temperature. However, oxidation of the dust collected from tile surfaces has led to ~ 18% mass loss after 8 hours; thereafter, little change in mass was observed. This suggests that the surface dust includes some components of different composition and/or structure – possibly fragments of codeposited layers. The oxidation of codeposit flakes scraped form DIII-D upper divertor tiles showed an initial 25% loss in mass due to heating in vacuum, and the gradual loss of 30-38% mass during the subsequent 24 hours exposure to O2. This behavior is significantly different from that observed for the oxidation of thinner DIII-D codeposit specimens which were still adhered to tile surfaces, and this is thought to be related to the low deuterium content (D/C ~ 0.03 – 0.04) of the flakes.

  4. Nonlinear fishbone dynamics in spherical tokamaks

    SciTech Connect

    Wang, Feng; Fu, G. Y.; Shen, Wei

    2016-11-22

    Linear and nonlinear kinetic-MHD hybrid simulations have been carried out to investigate linear stability and nonlinear dynamics of beam-driven fishbone instability in spherical tokamak plasmas. Realistic NSTX parameters with finite toroidal rotation were used. Our results show that the fishbone is driven by both trapped and passing particles. The instability drive of passing particles is comparable to that of trapped particles in the linear regime. The effects of rotation are destabilizing and a new region of instability appears at higher q min (>1.5) values, q min being the minimum of safety factor profile. In the nonlinear regime, the mode saturates due to flattening of beam ion distribution, and this persists after initial saturation while mode frequency chirps down in such a way that the resonant trapped particles move out radially and keep in resonance with the mode. Correspondingly, the flattening region of beam ion distribution expands radially outward. Furthermore, a substantial fraction of initially non-resonant trapped particles become resonant around the time of mode saturation and keep in resonance with the mode as frequency chirps down. On the other hand, the fraction of resonant passing particles is significantly smaller than that of trapped particles. Finally, our analysis shows that trapped particles provide the main drive to the mode in the nonlinear regime.

  5. Nonlinear fishbone dynamics in spherical tokamaks

    DOE PAGES

    Wang, Feng; Fu, G. Y.; Shen, Wei

    2016-11-22

    Linear and nonlinear kinetic-MHD hybrid simulations have been carried out to investigate linear stability and nonlinear dynamics of beam-driven fishbone instability in spherical tokamak plasmas. Realistic NSTX parameters with finite toroidal rotation were used. Our results show that the fishbone is driven by both trapped and passing particles. The instability drive of passing particles is comparable to that of trapped particles in the linear regime. The effects of rotation are destabilizing and a new region of instability appears at higher q min (>1.5) values, q min being the minimum of safety factor profile. In the nonlinear regime, the mode saturatesmore » due to flattening of beam ion distribution, and this persists after initial saturation while mode frequency chirps down in such a way that the resonant trapped particles move out radially and keep in resonance with the mode. Correspondingly, the flattening region of beam ion distribution expands radially outward. Furthermore, a substantial fraction of initially non-resonant trapped particles become resonant around the time of mode saturation and keep in resonance with the mode as frequency chirps down. On the other hand, the fraction of resonant passing particles is significantly smaller than that of trapped particles. Finally, our analysis shows that trapped particles provide the main drive to the mode in the nonlinear regime.« less

  6. Divertor design for the Tokamak Physics Experiment

    SciTech Connect

    Hill, D.N.; Braams, B.; Brooks, J.N.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4{times} L-mode), high beta ({beta}{sub N} {ge} 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74{degrees} from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m{sup 2} with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

  7. Tearing mode analysis in tokamaks, revisited

    SciTech Connect

    Nishimura, Y.; Callen, J.D.; Hegna, C.C.

    1997-12-01

    A new {Delta}{prime} shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio ({epsilon} {le} 0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of finite pressure term. Numerical results compare favorably with Furth et al. results. The effects of finite pressure, which are shown to decrease {Delta}{prime}, are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric element stabilizes the tearing mode significantly, even in a low {beta} regime before the toroidal magnetic curvature effects come into play. Double tearing modes in toroidal geometries are examined as well. Furthermore, m {ge} 2 tearing mode stability criteria are compared with three dimensional initial value MHD simulation by the FAR code.

  8. Zonal flows in tokamaks with anisotropic pressure

    NASA Astrophysics Data System (ADS)

    Ren, Haijun

    2014-04-01

    Zonal flows (ZFs) in a tokamak plasma with anisotropic pressure are investigated. The dynamics of perpendicular and parallel pressures are determined by the Chew-Goldberger-Low double equations and low-β condition is adopted, where β is the ratio of plasma pressure to the magnetic field pressure. The dispersion relation is analytically derived and illustrates two branches of ZFs. The low frequency zonal flow (LFZF) branch becomes unstable when χ, the ratio of the perpendicular pressure to the parallel one, is greater than a threshold value χc, which is about 3.8. In the stable region, its frequency increases first and then decreases with increasing χ. For χ = 1, the frequency of LFZF agrees well with the experimental observation. For the instability, the growth rate of LFZF increases with χ. The geodesic acoustic mode branch is shown to be always stable with a frequency increasing with χ. The safety factor is shown to diminish the frequencies of both branches or the growth rate of LFZF.

  9. Construction of the Lithium Tokamak Experiment (LTX)

    NASA Astrophysics Data System (ADS)

    Kozub, Thomas; Majeski, Richard; Kaita, Robert; Berzak, Laura; Lundberg, Daniel; Strickler, Trevor; Woolley, Robert; Zakharov, Leonid

    2008-11-01

    The Lithium Tokamak eXperiment (LTX)* will investigate the low recycling operating regime for magnetically confined plasmas using liquid lithium plasma facing surfaces. The engineering design and machine fabrication process will be presented. The most significant new feature of the LTX machine is the installation of a heated copper toroidal shell that will be operated at 300 C to 500 C. Its stainless steel plasma-facing liner will be internally coated with an evaporated layer of liquid lithium. The shell is comprised of four quadrants that have been fabricated in-house from explosively bonded stainless steel on copper to conform closely to the outer plasma flux surface. All internal components of the LTX machine have been designed and built to meet the simultaneous requirements for liquid lithium compatibility, high temperature operation, and electrical isolation. These requirements have led to unique design features, such as the method of supporting the shell quadrants, and construction of the new internal poloidal field coils. *Supported by US DOE contract #DE-AC02-76CH-03073

  10. Nonlinear fishbone dynamics in spherical tokamaks

    NASA Astrophysics Data System (ADS)

    Wang, Feng; Fu, G. Y.; Shen, Wei

    2017-01-01

    Linear and nonlinear kinetic-MHD hybrid simulations have been carried out to investigate linear stability and nonlinear dynamics of beam-driven fishbone instability in spherical tokamak plasmas. Realistic NSTX parameters with finite toroidal rotation were used. The results show that the fishbone is driven by both trapped and passing particles. The instability drive of passing particles is comparable to that of trapped particles in the linear regime. The effects of rotation are destabilizing and a new region of instability appears at higher q min (>1.5) values, q min being the minimum of safety factor profile. In the nonlinear regime, the mode saturates due to flattening of beam ion distribution, and this persists after initial saturation while mode frequency chirps down in such a way that the resonant trapped particles move out radially and keep in resonance with the mode. Correspondingly, the flattening region of beam ion distribution expands radially outward. A substantial fraction of initially non-resonant trapped particles become resonant around the time of mode saturation and keep in resonance with the mode as frequency chirps down. On the other hand, the fraction of resonant passing particles is significantly smaller than that of trapped particles. Our analysis shows that trapped particles provide the main drive to the mode in the nonlinear regime.

  11. ELM experimental study on the EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Zixi; Gao, Xiang; Xu, Xueqiao; Li, Jiangang; EAST Team

    2013-10-01

    Atypical Type III ELM is observed on EAST tokamak. This type of ELM has MHD precursor and high collisionality at the edge, and also the threshold power is close to the scaling law. But the frequency of the ELM does not decrease with the injected power. Power threshold is lower with the molybdenum wall in double null (DN) on EAST. Considering the effects of the plasma surface (S) to the threshold power, Double Null has the lowest power threshold. Better energy confinement has been observed in DN compared to Single-null (SN) at same power loss. But with the same power loss, Upper Single Null (USN) with the grad-B drift pointing backwards the active X-point (favorable direction) on EAST has the lower energy confinement time than Lower Single Null (LSN). Low Hybrid Wave (LHW) can mitigate ELMs. The power deposition should be near the edge in the H-mode phase. Not only the LHW decreases the max gradient of the density in the pedestal region, but also brings the density oscillations. Low X-point configurations in Lower single null have a lower power threshold. The low X-point discharges on EAST is closer to the DN. Approaching to the DN configuration should be the reason of the lower power threshold caused by the lower X-point on EAST.

  12. Ionization balance in EBIT and tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Peacock, N. J.; Barnsley, R.; O'Mullane, M. G.; Tarbutt, M. R.; Crosby, D.; Silver, J. D.; Rainnie, J. A.

    2001-01-01

    The equilibrium state in tokamak core plasmas has been studied using the relative intensities of resonance x-ray lines, for example Lyα (H-like), "w" (He-like), and "q" (Li-like) from test ions such as Ar+15, Ar+16, and Ar+17. A full spatial analysis involves comparison of the line intensities with ion diffusion calculations, including relevant atomic rates. A zero-dimensional model using a global ion loss rate approximation has also been demonstrated by comparison with the data collected from a Johann configuration spectrometer with a charged coupled device (CCD) detector. Since the lines are nearly monoenergetic, their intensities are independent of the instrument sensitivity and are directly proportional to the ion abundances. This method has recently been applied to Ar in the Oxford electron beam ion trap (EBIT) with a beam energy in the range 3-10 keV. Taking into account the cross sections for monoenergetic electron collisions and polarization effects, model calculations agree with the observed line ratios at 4.1 keV beam energy. This work will be expanded to provide nomograms of ionization state versus line intensity ratios as a function of EBIT beam energy.

  13. A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

    NASA Astrophysics Data System (ADS)

    Windsor, C. G.; Pautasso, G.; Tichmann, C.; Buttery, R. J.; Hender, T. C.; EFDA Contributors, JET; ASDEX Upgrade Team

    2005-05-01

    First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems.

  14. Maryland Controlled Fusion Research Program. Progress report, November 1, 1993--October 31, 1994

    SciTech Connect

    Antonsen, T.M. Jr.; Drake, J.F.; Finn, J.M.; Guzdar, P.N.; Hassam, A.; Liu, C.S.; Ott, E.

    1994-07-01

    The theoretical research activity at the University of Maryland supported by the Department of Energy focusses on two major tasks. They are: A. Transport, generation of shear flow and L-H transitions in Tokamaks; and B. MHD stability of Tokamaks. Titles of specific research projects under these fields carried out in this period include: Poloidal rotation of tokamak plasmas at super-poloidal-sonic speeds; Rotation of plasmas in neoclassical regimes; New unstable branch of drift resistive ballooning modes in tokamaks; 3D fluid simulations of the drift resistive ballooning modes - twisted coordinate system; Disintegration of ion banana orbits in tokamak edge plamsas; Fast reconnection in high temperature plasmas; Dynamics of sawtooth collapse in tokamak plasmas; Effect of trapped particles on MHD modes; MHD stability of high beta toroidal equilibria.

  15. Saturated internal instabilities in advanced-tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Hua, M.-D.; Chapman, I. T.; Pinches, S. D.; Hastie, R. J.; MAST Team

    2010-06-01

    "Advanced tokamak" (AT) scenarios were developed with the aim of reaching steady-state operation in future potential tokamak fusion power plants. AT scenarios exhibit non-monotonic to flat safety factor profiles (q, a measure of the magnetic field line pitch), with the minimum q (qmin) slightly above an integer value (qs). However, it has been predicted that these q profiles are unstable to ideal magnetohydrodynamic instabilities as qmin approaches qs. These ideal instabilities, observed and diagnosed as such for the first time in MAST plasmas with AT-like q profiles, have far-reaching consequences like confinement degradation, flattening of the toroidal core rotation or enhanced fast ion losses. These observations motivate the stability analysis of advanced-tokamak plasmas, with a view to provide guidance for stability thresholds in AT scenarios. Additionally, the measured rotation damping is compared to the self-consistently calculated predictions from neoclassical toroidal viscosity theory.

  16. Prospects and status of low-aspect-ratio tokamaks

    SciTech Connect

    Peng, Y.K.M.

    1994-12-31

    The prospects for the low-aspect-ratio (A) tokamak to fulfill the requirements of viable fusion power plants are considered relative to the present status in data and modeling. Desirable physics and design features for an attractive Blanket Test Facility and power reactors are estimated for low-A tokamaks based on calculations improved with the latest data from small pioneering experiments. While these experiments have confirmed some of the recent predictions for low-A, they also identify the remaining issues that require verification before reliable projections can be made for these deuterium-tritium applications. The results show that the low-A regime of small size, modest field, and high current offers a path complementary to the standard and high A tokamaks in developing the full potential of fusion power.

  17. Stochastic modeling of plasma mode forecasting in tokamak

    NASA Astrophysics Data System (ADS)

    Saadat, Sh.; Salem, M.; Ghoranneviss, M.; Khorshid, P.

    2012-04-01

    The structure of magnetohydrodynamic (MHD) modes has always been an interesting study in tokamaks. The mode number of tokamak plasma is the most important parameter, which plays a vital role in MHD instabilities. If it could be predicted, then the time of exerting external fields, such as feedback fields and Resonance Helical Field, could be obtained. Autoregressive Integrated Moving Average (ARIMA) and Seasonal Autoregressive Integrated Moving Average are useful models to predict stochastic processes. In this paper, we suggest using ARIMA model to forecast mode number. The ARIMA model shows correct mode number (m = 4) about 0.5 ms in IR-T1 tokamak and equations of Mirnov coil fluctuations are obtained. It is found that the recursive estimates of the ARIMA model parameters change as the plasma mode changes. A discriminator function has been proposed to determine plasma mode based on the recursive estimates of model parameters.

  18. Hybrid Method for Tokamak MHD Equilibrium Configuration Reconstruction

    NASA Astrophysics Data System (ADS)

    He, Hong-Da; Dong, Jia-Qi; Zhang, Jin-Hua; Jiang, Hai-Bin

    2007-02-01

    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.

  19. A method for determining poloidal coil configurations for tokamak devices

    SciTech Connect

    Evans, K. Jr.

    1990-12-01

    This paper presents a method for obtaining the locations and currents of the poloidal coil systems for a tokamak, given an desirable magnetohydrodynamic equilibrium for the device. The method involves a simultaneous minimization of the match to the desired poloidal field and the stored energy in the coils, subject to the constraints necessary to achieve decoupling of the equilibrium and inductive-current-drive (ohmic-heating) systems and to achieve a given coupling of the current-drive system with the plasma. A compendium of mutual and self-inductance formulas as they apply to tokamak systems is presented, as well as examples of how the method has been used in the design of several tokamaks. Finally, a user manual for a computer code that implements this method is provided. 14 refs., 11 figs., 1 tab.

  20. Runaway electrons in a tokamak: A free-electron maser

    SciTech Connect

    Kurzan, B.; Steuer, K.

    1997-04-01

    In ohmic divertor plasma discharges of the ASDEX upgrade tokamak containing a small population of runaway electrons, fluctuating emission in the microwave region with a very narrow bandwidth is observed. The radiation can be explained by relativistic runaway electrons, which are captured in a ripple resonance of the tokamak and are thus made monoenergetic enough that they can undergo the collective instability of a free-electron maser. From the frequency of the maser, the energy of the runaway electrons, and from the linewidth and energy per radiation pulse, the particle density of the runaway electrons is determined locally. Observing this maser radiation is thus a different diagnostic for runaway electrons in tokamaks. {copyright} {ital 1997} {ital The American Physical Society}

  1. The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

    PubMed Central

    Li, Jing; Quan, Tingting; Zhang, Wei; Deng, Wei

    2014-01-01

    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. PMID:24892099

  2. The dynamic mutation characteristics of thermonuclear reaction in Tokamak.

    PubMed

    Li, Jing; Quan, Tingting; Zhang, Wei; Deng, Wei

    2014-01-01

    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 Z 2-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.

  3. Fractal structure of films deposited in a tokamak

    SciTech Connect

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

    2007-04-15

    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 {mu}m 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.

  4. Resonant-cavity ICRF coupler for large tokamaks

    SciTech Connect

    Perkins, F.W.; Kluge, R.F.

    1983-04-01

    A new resonant-cavity ICRF coupler is proposed for large tokamaks. The design features a novel resonant cavity, an rf magnetic-field orientation that effectively radiates fast Alfven waves, matching to 40 ..cap omega.. transmission lines, and an electric-field orientation so that the strongest rf electric fields are orthogonal to the main toroidal magnetic field thereby benefitting from magnetic insulation. As a result, the power handling capability is excellent. For the case of the Big-Dee Doublet III tokamak, a single 35 cm x 50 cm coupler can launch 20 MW of fast Alfven waves. Extrapolation to fusion reactor parameters is straightforward.

  5. Rotational resonance of nonaxisymmetric magnetic braking in the KSTAR tokamak.

    PubMed

    Park, J-K; Jeon, Y M; Menard, J E; Ko, W H; Lee, S G; Bae, Y S; Joung, M; You, K-I; Lee, K-D; Logan, N; Kim, K; Ko, J S; Yoon, S W; Hahn, S H; Kim, J H; Kim, W C; Oh, Y-K; Kwak, J-G

    2013-08-30

    One of the important rotational resonances in nonaxisymmetric neoclassical transport has been experimentally validated in the KSTAR tokamak by applying highly nonresonant n=1 magnetic perturbations to rapidly rotating plasmas. These so-called bounce-harmonic resonances are expected to occur in the presence of magnetic braking perturbations when the toroidal rotation is fast enough to resonate with periodic parallel motions of trapped particles. The predicted and observed resonant peak along with the toroidal rotation implies that the toroidal rotation in tokamaks can be controlled naturally in favorable conditions to stability, using nonaxisymmetric magnetic perturbations.

  6. A Midsize Tokamak As Fast Track To Burning Plasmas

    SciTech Connect

    E. Mazzucato

    2010-07-14

    This paper presents 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 could 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 more efficient magnetic divertor than those of present tokamaks is discussed.

  7. Resistive wall mode stabilization by plasma rotation in advanced tokamaks

    NASA Astrophysics Data System (ADS)

    Eriksson, G.

    1996-03-01

    By combining previous results of Betti and Freidberg [Phys. Rev. Lett. 74, 2949 (1995)] and Eriksson [Phys. Plasmas 2, 3095 (1995)], a fully analytical description is obtained for the stabilizing effect of toroidal plasma rotation in a large aspect ratio tokamak surrounded by a resistive wall. As in advanced tokamak configurations with a large fraction of bootstrap current, it is assumed that the current gradient near the plasma edge is large. This assumption enables an analytical analysis of external kink modes with low poloidal mode numbers. An expression is obtained, showing explicitly how the window of stable wall distances depends on the current profile.

  8. Accessibility of second regions of stability in tokamaks

    SciTech Connect

    Manickam, J.

    1985-12-01

    Second regions of stability to the ideal ballooning modes have been shown to exist in large-aspect-ratio circular and small-aspect-ratio bean-shaped tokamaks. We report on the existence of these second stability regions in finite-aspect-ratio dee-shaped tokamaks. We also report on the discovery of a second-stable region with respect to the n = 1 external kink mode in a bean-shaped plasma. The role of the shear and current profile in determining these regions of parameter space are discussed. 13 refs., 6 figs.

  9. Resistive demountable toroidal-field coils for tokamak reactors

    SciTech Connect

    Jassby, D.L.; Jacobsen, R.A.; Kalnavarns, J.; Masson, L.S.; Sekot, J.P.

    1981-07-01

    Readily demountable TF (toroidal-field) coils allow complete access to the internal components of a tokamak reactor for maintenance of replacement. The requirement of readily demountable joints dictates the use of water-cooled resistive coils, which have a host of decisive advantages over superconducting coils. Previous papers have shown that resistive TF coils for tokamak reactors can operate in the steady state with acceptable power dissipation (typically, 175 to 300 MW). This paper summarizes results of parametric studies of size optimization of rectangular TF coils and of a finite-element stress analysis, and examines several candidate methods of implementing demountable joints for rectangular coils constructed of plate segments.

  10. The role of spherical torus in clarifying tokamak physics

    SciTech Connect

    Morris, A. W.; Peng, Yueng Kay Martin

    1999-01-01

    The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one s understanding will be emphasized.

  11. Kinetic Energy Principle And Neoclassical Toroidal Torque In Tokamaks

    SciTech Connect

    Jong-Kyu Park

    2011-11-07

    It is shown that when tokamaks are perturbed the kinetic energy principle is closely related to the neoclassical toroidal torque by the action invariance of particles. Especially when tokamaks are perturbed from scalar pressure equilibria, the imaginary part of the potential energy in the kinetic energy principle is equivalent to the toroidal torque by the Neoclassical Toroidal Viscosity (NTV). A unified description therefore should be made for both physics. It is also shown in this case that the potential energy operator can be self-adjoint and thus the stability calculation can be simplified by minimizing the potential energy

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

    SciTech Connect

    Kuprat, A.P.; Glasser, A.H.

    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.

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

    DOE PAGES

    Liu, Jian; Qin, Hong; Fisch, Nathaniel J.; ...

    2014-06-19

    In this study, 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.

  14. Internal Magnetic Configuration Measured by ECE Imaging on EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Xu, Ming; Wen, Yizhi; Xie, Jinlin; Yu, Changxuan; Gao, Bingxi; Xu, Xiaoyuan; Liu, Wandong; Hu, Liqun; Sun, Youwen; Qian, Jinping; Wan, Baonian

    2013-12-01

    ECE imaging (electron cyclotron emission imaging) is an important diagnostic which can give 2D imaging of temperature fluctuation in the core of tokamak. A method based on ECE imaging is introduced which can give the information of the position of magnetic axis and the structure of internal magnetic surface for EAST tokamak. The EFIT equilibrium reconstruction is not reliable due to the absence of important core diagnostic at the initial phase for EAST, so the information given by ECE imaging could help to improve the accuracy of EFIT equilibrium reconstruction.

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

    SciTech Connect

    Hively, L.M.

    1980-01-01

    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-..beta.., non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated.

  16. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    SciTech Connect

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  17. Magnetic flux reconstruction methods for shaped tokamaks

    NASA Astrophysics Data System (ADS)

    Tsui, Chi-Wa

    1993-12-01

    The use of a variational method permits the Grad-Shafranov (GS) equation to be solved by reducing the problem of solving the two dimensional nonlinear partial differential equation to the problem of minimizing a function of several variables. This high speed algorithm approximately solves the GS equation given a parameterization of the plasma boundary and the current profile (p' and FF' functions). The current profile parameters are treated as unknowns. The goal is to reconstruct the internal magnetic flux surfaces of a tokamak plasma and the toroidal current density profile from the external magnetic measurements. This is a classic problem of inverse equilibrium determination. The current profile parameters can be evaluated by several different matching procedures. Matching of magnetic flux and field at the probe locations using the Biot-Savart law and magnetic Green's function provides a robust method of magnetic reconstruction. The matching of poloidal magnetic field on the plasma surface provides a unique method of identifying the plasma current profile. However, the power of this method is greatly compromised by the experimental errors of the magnetic signals. The Casing principle provides a very fast way to evaluate the plasma contribution to the magnetic signals. It has the potential of being a fast matching method. The performance of this method is hindered by the accuracy of the poloidal magnetic field computed from the equilibrium solver. A flux reconstruction package has been implemented which integrates a vacuum field solver using a filament model for the plasma, a multilayer perception neural network as an interface, and the volume integration of plasma current density using Green's functions as a matching method for the current profile parameters. The flux reconstruction package is applied to compare with the ASEQ and EFIT data.

  18. Magnetic flux reconstruction methods for shaped tokamaks

    SciTech Connect

    Tsui, Chi-Wa

    1993-12-01

    The use of a variational method permits the Grad-Shafranov (GS) equation to be solved by reducing the problem of solving the 2D non-linear partial differential equation to the problem of minimizing a function of several variables. This high speed algorithm approximately solves the GS equation given a parameterization of the plasma boundary and the current profile (p` and FF` functions). The author treats the current profile parameters as unknowns. The goal is to reconstruct the internal magnetic flux surfaces of a tokamak plasma and the toroidal current density profile from the external magnetic measurements. This is a classic problem of inverse equilibrium determination. The current profile parameters can be evaluated by several different matching procedures. Matching of magnetic flux and field at the probe locations using the Biot-Savart law and magnetic Green`s function provides a robust method of magnetic reconstruction. The matching of poloidal magnetic field on the plasma surface provides a unique method of identifying the plasma current profile. However, the power of this method is greatly compromised by the experimental errors of the magnetic signals. The Casing Principle provides a very fast way to evaluate the plasma contribution to the magnetic signals. It has the potential of being a fast matching method. The performance of this method is hindered by the accuracy of the poloidal magnetic field computed from the equilibrium solver. A flux reconstruction package has been implemented which integrates a vacuum field solver using a filament model for the plasma, a multi-layer perception neural network as an interface, and the volume integration of plasma current density using Green`s functions as a matching method for the current profile parameters. The flux reconstruction package is applied to compare with the ASEQ and EFIT data. The results are promising.

  19. Far infrared tangential interferometry/polarimetry on the National Spherical Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    Park, H. K.; Domier, C. W.; Geck, W. R.; Luhmann, N. C.

    1999-01-01

    Measurement of the core BT(r,t) value is essential in the National Spherical Tokamak Experiment (NSTX), since the effects of paramagnetism and diamagnetism in the NSTX are expected to be considerably greater than that in higher aspect ratio tokamaks. Therefore, without independent BT(r,t) measurement, plasma parameters dependent upon BT such as the q profile and local β value cannot be evaluated. Tangential interferometer/polarimeter systems (eight channels) [H. Park, L. Guttadora, C. Domier, W. R. Geck, and N. C. Luhman, Jr., First and Second NSTX Research Forums, Princeton, NJ, 1997 (unpublished)] for the NSTX will provide temporally and radially resolved toroidal field profile [BT(r,t)] and two-dimensional electron density profile [ne(r,t)] data. The outcome of the proposed system is extremely important to the study of confinement, heating, and stability of the NSTX plasmas. The research task is largely based on utilizing existing hardware from the TFTR multichannel infrared interferometer system [D. K. Mansfield, H. K. Park, L. C. Johnson, H. Anderson, S. Foote, B. Clifton, and C. H. Ma, Appl. Opt. 26, 4469 (1987) and H. K. Park, D. K. Mansfield, and C. L. Johnson, Proceedings of the 3rd International Symposium on Laser-Aided Plasma Diagnostic, Los Angeles, CA, 28-30 Oct. 1987 (unpublished), pp. 96-104] which will be reconfigured into a tangential system for NSTX, and to develop the additional hardware required to complete the system.

  20. Analysis of images from videocameras in the Frascati Tokamak Upgrade tokamak

    SciTech Connect

    De Angelis, R.; Migliori, S.; Borioni, S.; Bracco, G.; Pierattini, S.; Perozziello, A.

    2004-10-01

    The plasma edge interaction in FTU tokamak is monitored by wide angle videocameras. Data are acquired as movies or single frames at a rate of 50 frames/s. The images show interesting features of the plasma such as the presence of Marfes or runaways and give useful information on the status of large parts of the vacuum vessel and toroidal limiter. Due to the large number of data available visual inspection of the movies is often insufficient to correlate the images to the experimental findings. This article illustrates a number of applications developed in order to correlate the images with plasma signals and to search the image database for specific features relevant to the discharge.

  1. HPGe well-type detectors for neutron activation measurements on the Frascati Tokamak Upgrade tokamak

    SciTech Connect

    Bertalot, L.; Damiani, M.; Esposito, B.; Lagamba, L.; Podda, S.; Batistoni, P.; De Felice, P.; Biagini, R.

    1997-01-01

    We describe an improvement of the neutron activation system in operation on the Frascati Tokamak Upgrade (FTU) tokamak for the measurement of the total neutron yield. A HPGe well-type detector (200 cm{sup 3} active volume) is used to detect the photoemission from neutron activated samples ({sup 115m}In336.2 keV {gamma} rays from DD neutrons on indium for FTU). Due to their high geometrical efficiency, HPGe well-type detectors are particularly suited to the FTU low-level activity measurements. A particular effort has been devoted to the calibration of the measuring system. In particular, a multi-{gamma} calibration source (59{endash}1332 keV energy range) with a density of 7.31 g/cm{sup 3} consisting of a stack of indium foils has been prepared. This assures that the shape and volume of the calibration source are the same as those of the samples used in the actual measurements. The full-energy-peak efficiency at the {sup 115m}In336.2 keV line is 0.197 with an overall uncertainty of 2{percent} (1{sigma}). For a better characterization of the detector response as a function of the sample density, a further calibration source with the same geometry has been prepared in a gel aqueous solution (density {approximately}1 g/cm{sup 3}). The calibration curves for the well-type detector at the two different density values are compared. {copyright} {ital 1997 American Institute of Physics.}

  2. Midplane Faraday rotation: A densitometer for large tokamaks

    NASA Astrophysics Data System (ADS)

    Jobes, F. C.; Mansfield, D. K.

    1992-10-01

    The density in a large tokamak such as International Thermonuclear Experimental Reactor (ITER), or any of the proposed future US machines, can be determined by measuring the Faraday rotation of a 10.6 μm laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then ne(R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, operated at a full field of 30 T m and a density of 2×1020/m3, the rotation angle would be quite large-about 60° for two passes. A layout in which a single laser beam is fanned out in the horizontal midplane of the tokamak, with a set of retroreflectors on the far side of the vacuum vessel, would provide good spatial resolution, depending only upon the number of reflectors. With this proposed layout, only one window would be needed. Because the rotation angle is never more than 1 ``fringe,'' the data is always good, and it is also a continuous measurement in time. Faraday rotation is dependent only upon the plasma itself, and thus is not sensitive to vibration of the optical components. Simulations of the expected results show that ITER, or any large tokamak, existing or proposed, would be well served even at low densities by a midplane Faraday rotation densitometer of ˜64 channels.

  3. Gamma ray imager on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    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.

  4. Observation of finite-. beta. MHD phenomena in tokamaks

    SciTech Connect

    McGuire, K.M.

    1984-09-01

    Stable high-beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1 internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of q/sub 0/ and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas, two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded. Whether this represents an ultimate tokamak limit or if beta optimized configurations (Dee- or bean-shaped plasmas) can exceed this limit and perhaps enter a second regime of stability remains to be clarified.

  5. Electron-cyclotron-heating experiments in tokamaks and stellarators

    SciTech Connect

    England, A.C.

    1983-01-01

    This paper reviews the application of high-frequency microwave radiation to plasma heating near the electron-cyclotron frequency in tokamaks and stellarators. Successful plasma heating by microwave power has been demonstrated in numerous experiments. Predicted future technological developments and current theoretical understanding suggest that a vigorous program in plasma heating will continue to yield promising results.

  6. TPX diagnostics for tokamak operation, plasma control and machine protection

    SciTech Connect

    Edmonds, P.H.; Medley, S.S.; Young, K.M.

    1995-08-01

    The diagnostics for TPX are at an early design phase, with emphasis on the diagnostic access interface with the major tokamak components. Account has to be taken of the very severe environment for diagnostic components located inside the vacuum vessel. The placement of subcontracts for the design and fabrication of the diagnostic systems is in process.

  7. Tight aspect ratio tokamak experiments and prospects for the future

    SciTech Connect

    Sykes, A; Peng, Yueng Kay Martin

    1995-01-01

    The present status of experimental results from low aspect ratio tokamaks is described, together with plans for physics experiments at the mega-amp level. Further development of the concept, and its potential for a materials/component test facility or ultimately a fusion power plant, are indicated.

  8. Ballooning mode stability of elongated high-beta tokamaks

    NASA Astrophysics Data System (ADS)

    Mauel, Michael E.

    1987-12-01

    The variational principle derived by Choe and Freidberg [Phys. Fluids 29, 1766 (1986)] and used to estimate the geometry of high-beta tokamak equilibria is extended to include elongation. Ballooning mode stability is then investigated, illustrating the influence of elongation on local and global stability.

  9. Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

    ScienceCinema

    None

    2016-07-12

    From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

  10. Dynamic diagnostics of the error fields in tokamaks

    NASA Astrophysics Data System (ADS)

    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.

  11. INTOR: a first-generation tokamak experimental reactor

    SciTech Connect

    Stacey, Jr, W M; Gilleland, J R; Kulcinski, G L; Rutherford, P H

    1980-02-01

    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.

  12. First neutron spectrometry measurement at the HL-2A Tokamak

    NASA Astrophysics Data System (ADS)

    Yuan, Xi; Zhang, Xing; Xie, Xu-Fei; Chen, Zhong-Jing; Peng, Xing-Yu; Fan, Tie-Shuan; Chen, Jin-Xiang; Li, Xiang-Qing; Yuan, Guo-Liang; Yang, Qing-Wei; Yang, Jin-Wei

    2013-12-01

    A compact neutron spectrometer based on the liquid scintillator is presented for neutron energy spectrum measurements at the HL-2A Tokamak. The spectrometer was well characterized and a fast digital pulse shape discrimination software was developed using the charge comparison method. A digitizer data acquisition system with a maximum frequency of 1 MHz can work under an environment with a high count rate at HL-2A Tokamak. Specific radiation and magnetic shielding for the spectrometer were designed for the neutron spectrum measurement at the HL-2A Tokamak. For pulse height spectrum analysis, dedicated numerical simulation utilizing NUBEAM combined with GENESIS was performed to obtain the neutron energy spectrum. Subsequently, the transportation process from the plasma to the detector was evaluated with Monte Carlo calculations. The distorted neutron energy spectrum was folded with the response matrix of the liquid scintillation spectrometer, and good consistency was found between the simulated and measured pulse height spectra. This neutron spectrometer based on a digital acquisition system could be well adopted for the investigation of the auxiliary heating behavior and the fast-ion related phenomenon on different tokamak devices.

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

    PubMed

    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.

  14. Induced emission of extraordinary mode radiation in tokamaks

    NASA Technical Reports Server (NTRS)

    Freund, H. P.; Lee, L. C.

    1979-01-01

    The implications of the formation of a positive slope in the runaway electron tail in tokamak plasmas are investigated in regard to the radiation in the vicinity of the electron plasma frequency. In particular, it is shown that the amplification of extraordinary mode waves may result.

  15. Test results on systems developed for the SST-1 tokamak

    NASA Astrophysics Data System (ADS)

    Bora, D.; SST-1 Team

    2003-12-01

    The steady state superconducting tokamak (SST-1) is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation (kgr) and triangularity (dgr). Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. A NbTi based cable-in-conduit conductor (CICC) has been fabricated by M/S Hitachi Cables Ltd., Japan under specification and supervision of the Institute for Plasma Research (IPR). The suitability of this CICC for the SST-1 magnets has been validated through test carried out on a model coil wound from this CICC. Toroidal and poloidal SC magnets have been fabricated and factory acceptance tests have been performed. SC magnets require liquid helium (LHe) cooled current leads, electrical isolators at LHe temperature, superconducting bus bars and LHe transfer lines. Full scale prototypes of these have been developed and tested successfully. SC magnets will be cooled to 4.5 K by forced flow of supercritical helium through the CICC. A 1 kW grade liquefier/refrigerator has been installed and is in final stages of commissioning at IPR. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors (VSs) having ports and 16 rings with {\\sf D} -shaped cross-section. To establish the fabrication methodology for this, a fullscale prototype of the vessel with two VSs and three rings has been fabricated and tested successfully. Based on this the fabrication of the VSs and rings is in final stage of fabrication. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. SST-1 will have three different high power radio frequency systems to additionally heat and non-inductively drive plasma current to sustain the plasma in steady state for a duration of up to 1000 s. Ion cyclotron resonance frequency (ICRF) and electron cyclotron resonance frequency (ECRF

  16. UCLA program in reactor studies: The ARIES tokamak reactor study. Progress report, December 1, 1990--November 30, 1991

    SciTech Connect

    Not Available

    1991-12-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. 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. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ``modest`` extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  17. LIDAR Thomson scattering for advanced tokamaks. Final report

    SciTech Connect

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

    1996-03-18

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured.

  18. Analytic model for coaxial helicity injection in tokamak plasmas

    SciTech Connect

    Weening, R. H.

    2011-12-15

    Using a partial differential equation for the time evolution of the mean-field poloidal magnetic flux that incorporates resistivity {eta} and hyper-resistivity {Lambda} terms, an exact analytic solution is obtained for steady-state coaxial helicity injection (CHI) in force-free large aspect ratio tokamaks. The analytic mean-field Ohm's law model allows for calculation of the tokamak CHI current drive efficiency and the plasma inductances at arbitrary levels of magnetic fluctuations, or dynamo activity. The results of the mean-field model suggest that CHI approaching Ohmic efficiency is only possible in tokamaks when the size of the effective current drive boundary layer, {delta}{identical_to}({Lambda}/{eta}){sup 1/2}, becomes greater than half the size of the plasma, {delta}>a/2, with a the plasma minor radius. The electron thermal diffusivity due to magnetic fluctuation induced transport is obtained from the expression {chi}{sub e}={Lambda}/{mu}{sub 0}d{sub e}{sup 2}, with {mu}{sub 0} the permeability of free space and d{sub e} the electron skin depth, which for typical tokamak fusion plasma parameters is on the order of a millimeter. Thus, the ratio of the energy confinement time to the resistive diffusion time in a tokamak plasma driven by steady-state CHI approaching Ohmic efficiency is shown to be constrained by the relation {tau}{sub E}/{tau}{sub {eta}}<(d{sub e}/a){sup 2}{approx_equal}10{sup -6}. The mean-field model suggests that steady-state CHI can be viewed most simply as a boundary layer of stochastically wandering magnetic field lines.

  19. A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak

    NASA Astrophysics Data System (ADS)

    Ren, J.; Zuo, G. Z.; Hu, J. S.; Sun, Z.; Yang, Q. X.; Li, J. G.; Zakharov, L. E.; Xie, H.; Chen, Z. X.

    2015-02-01

    A program involving the extensive and systematic use of lithium (Li) as a "first," or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.

  20. Extreme ultraviolet diagnostic upgrades for kink mode control on the HBT-EP tokamak

    NASA Astrophysics Data System (ADS)

    Levesque, J. P.; Brooks, J. W.; Desanto, S.; Mauel, M. E.; Navratil, G. A.; Page, J. W.; Hansen, C. J.; Delgado-Aparicio, L.

    2016-10-01

    Optical diagnostics can provide non-invasive measurements of tokamak equilibria and the internal characteristics of MHD mode activity. We present research plans and ongoing progress on upgrading extreme ultraviolet (EUV) diagnostics in the HBT-EP tokamak. Four sets of 16 poloidal views will allow tomographic reconstruction of plasma emissivity and internal kink mode structure. Emission characteristics of naturally-occurring m/n = 2/1, 3/2, and 3/1 tearing and kink modes will be compared with expectations from a synthetic diagnostic. Coupling between internal and external modes leading up to disruptions is studied. The internal plasma response to external magnetic perturbations is investigated, and compared with magnetic response measurements. Correlation between internal emissivity and external magnetic measurements provides a global picture of long-wavelength MHD instabilities. Measurements are input to HBT-EP's GPU-based feedback system, allowing active feedback for kink modes using only optical sensors and both magnetic and edge current actuators. A separate two-color, 16-chord tangential system will be installed next year to allow reconstruction of temperature profiles and their fluctuations versus time. Supported by U.S. DOE Grant DE-FG02-86ER53222.

  1. CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES

    SciTech Connect

    STRAIT,EJ; BIALEK,J; CHANCE,MS; CHU,MS; EDGELL,DH; FERRON,JR; GREENFIELD,CM; GAROFALO,AM; HUMPHREYS,DA; JACKSON,GL; JAYAKUMAR,RJ; JERNIGAN,TC; KIM,JS; LA HAYE,RJ; LAO,LL; LUCE,TC; MAKOWSKI,MA; MURAKAMI,M; NAVRATIL,GA; OKABAYASHI,M; PETTY,CC; REIMERDES,H; SCOVILLE,JT; TURNBULL,AD; WADE,MR; WALKER,ML; WHYTE,DG; DIII-D TEAM

    2003-06-01

    OAK-B135 Advanced tokamak research in DIII-D seeks to optimize the tokamak approach for fusion energy production, leading to a compact, steady state power source. High power density implies operation at high toroidal beta, {beta}{sub T}=

    2{micro}{sub 0}/B{sub T}{sup 2}, since fusion power density increases roughly as the square of the plasma pressure. Steady-state operation with low recirculating power for current drive implies operation at high poloidal beta, {beta}{sub P} =

    2{micro}{sub 0}/{sup 2}, in order to maximize the fraction of self-generated bootstrap current. Together, these lead to a requirement of operation at high normalized beta, {beta}{sub N} = {beta}{sub T}(aB/I), since {beta}{sub P}{beta}{sub T} {approx} 25[(1+{kappa}{sup 2})/2] ({beta}{sub N}/100){sup 2}. Plasmas with high normalized beta are likely to operate near one or more stability limits, so control of MHD stability in such plasmas is crucial.

  2. Real-time extraction of plasma equilibrium parameters in KSTAR tokamak using statistical methods

    NASA Astrophysics Data System (ADS)

    Na, Yong-Su; Jeon, Young-Mu; Hong, S. H.; Hwang, Y. S.

    2001-02-01

    To improve inherent shortcomings of statistical methods and apply them to the extraction of plasma equilibrium parameters in a fast timescale for real-time plasma control, new concepts of statistical methods such as principal component analysis-based neural network (NN), functional parametrization (FP)-based NN and double network are introduced by modifying NN and FP. These new methods are benchmarked and compared with the conventional techniques of NN and FP in a simple single-filament system. As a result of their applications to identification of plasma equilibrium parameters in the Korea Superconducting Tokamak Advanced Research tokamak, particularly, the double network concept among them has successfully achieved the improvement of drawbacks in the conventional methods. It is shown that more reliable results from the double network method can be obtained by combining several different statistical treatments as a primary network. Even in the case of nonoptimized methods united as a primary network, quite acceptable results can be achieved in the double network method.

  3. Geodesic acoustic modes in toroidally rotating tokamaks with an arbitrary β

    SciTech Connect

    Ren, Haijun; Li, Ding; Chu, Paul K

    2013-07-15

    Theoretical research on the geodesic acoustic mode (GAM) induced by the equilibrium toroidal rotation flow (ETRF) in the tokamak plasmas with an arbitrary β is performed by using the ideal magnetohydrodynamic model, where β is the ratio of the plasma pressure and magnetic field pressure. Two equations determining the poloidal displacement ξ{sub θ} and the divergence of the Lagrangian perturbation are obtained and suitable for arbitrary cross-section tokamaks with large-aspect-ratios. The dispersion relations are then derived for two different coupling patterns by assuming ξ{sub ±2}=0 and ξ{sub ±4}=0, respectively, where ξ{sub m}=(1/2π)∫ξ{sub θ}e{sup imθ}dθ with θ being the poloidal angle under the circular cross-section condition. In both patterns, the ETRF will increase the frequencies of the GAMs but β can decrease them. The GAM for ξ{sub ±2}=0 has a larger frequency than GAM for ξ{sub ±4}=0.

  4. Preliminary Design of the Alfvén Antennas on the J-TEXT Tokamak

    NASA Astrophysics Data System (ADS)

    He, Jiyang; Hu, Qiming; Rao, Bo; Liu, Linzi; Zhuang, Ge; J-TEXT Team

    2015-11-01

    Research on Alfvén waves and Alfvén eigenmodes(AEs) is of importance in tokamak plasma physics, such as investigation of interaction between energetic particles and AEs, turbulence and anomalous transport due to AEs, and so on. In order to study the Alfvén eigenmode excitation, damping features and the interaction between AEs and plasma transport, an Alfvén antennas system is designed for the J-TEXT tokamak. The system can generate high frequency magnetic field aiming to excite the AEs, especially toroidal Alfvén eigenmodes (TAE). For a typical J-TEXT plasma (BT = 1-2.2 T,ne = 3-6 ×1019m-3), the computed gaps in the Alfvén continua range from 300 to 500 kHz, with respect to the Alfvén waves dispersion relation. Three pairs of antennas are designed at different toroidal angles respectively on the low field side. Each pair consisting of two coils installed with angles of +/- 45° off the mid-plane along the poloidal direction. With this system, magnetic field components of mode number m =1-10, n =1-20 can be produced. The calculations show that the magnetic field in the LCFS can reach ~ 10Gs totally while about ~ 0.5Gs for each mode number, with the coil current of 20A. Corresponding author

  5. Physics of radiation-driven islands near the tokamak density limit

    NASA Astrophysics Data System (ADS)

    Gates, D. A.; Delgado-Aparicio, L.; White, R. B.

    2013-06-01

    In previous work (Gates and Delgado-Aparicio 2012 Phys. Rev. Lett. 108 165004), the onset criterion for radiation-driven islands (Rebut et al 1985 Proc. 10th Int. Conf. on Plasma Physics and Controlled Nuclear Fusion Research 1984 (London, UK, 1984) vol 2 (Vienna: IAEA) p 197) in combination with a simple cylindrical model of tokamak current channel behaviour was shown to be consistent with the empirical scaling of the tokamak density limit (Greenwald et al 1988 Nucl. Fusion 28 2199). A number of the unexplained phenomena at the density limit are consistent with this novel physics mechanism. In this work, a more formal theoretical underpinning, consistent with cylindrical tearing mode theory, is developed for the onset criteria of these modes. The appropriate derivation of the radiation-driven addition to the modified Rutherford equation (MRE) is discussed. Additionally, the ordering of the terms in the MRE is examined in a regime near the density limit. It is hoped that, given the apparent success of this simple model in explaining the observed global scalings, it will lead to a more comprehensive analysis of the possibility that radiation-driven islands are the physics mechanism responsible for the density limit. In particular, with modern diagnostic capabilities detailed measurements of current densities, electron densities and impurity concentrations at rational surfaces should be possible, enabling verification of the concepts described above.

  6. Non-linear MHD Simulation of ELMs including Pellet Triggered ones for KSTAR tokamak

    NASA Astrophysics Data System (ADS)

    Han, Hyunsun; Park, G.; Strauss, H.; Kim, J. Y.

    2011-10-01

    Three-dimensional non-linear MHD simulations have been conducted to investigate the qualitative characteristics of ELM(Edge Localized Mode)s including pellet induced ones using the M3D code. A linearized velocity perturbation of initial equilibrium is employed to trigger the ELM instability for the simulation of natural ELM, while a density blob, which represents the ionized pellet ablation and is located within the edge pedestal, is adopted in an adiabatic condition for that of pellet induced one. The initial equilibrium is constructed based on a H-mode plasma of KSTAR(Korea Superconducting Tokamak Advanced Research) device. It is found that characteristics of natural ELM simulation are in qualitative agreement with the experimental observations including that density perturbation is much larger than temperature one during ELM instability. Regarding the pellet induced ELM, it is observed that the locally increased pressure due to the fast parallel heat conduction compared to the spread of density perturbation triggers the peeling-ballooning instability resulting in ELM-like relaxation. Detailed results will be presented in the discussion of underlying mechanism and application to KSTAR tokamak.

  7. A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak

    SciTech Connect

    Ren, J.; Zuo, G. Z.; Hu, J. S.; Sun, Z.; Yang, Q. X.; Li, J. G.; Xie, H.; Chen, Z. X.; Zakharov, L. E.

    2015-02-15

    A program involving the extensive and systematic use of lithium (Li) as a “first,” or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.

  8. Non-axisymmetric equilibrium reconstruction for stellarators, reversed field pinches and tokamaks

    SciTech Connect

    Hanson, James D.; Anderson, D.T.; Cianciosa, M.; Franz, P.; Hartwell, G. H.; Hirshman, Steven Paul; Knowlton, Stephen F.; Lao, Lang L.; Lazarus, Edward Alan; Marrelli, L.; Maurer, D. A.; Schmitt, J. C.; Sontag, A. C.; Stevenson, B. A.; Terranova, D.

    2013-01-01

    Axisymmetric equilibrium reconstruction using magnetohydrodynamic equilibrium solutions to the Grad Shafranov equation has long been an important tool for interpreting tokamak experiments. This paper describes recent results in non-axisymmetric (three-dimensional) equilibrium reconstruction of nominally axisymmetric plasmas (tokamaks and reversed field pinches (RFPs)), and fully non-axisymmetric plasmas (stellarators). Results from applying the V3FIT code to CTH and HSX stellarator plasmas, RFX-mod RFP plasmas and the DIII-D tokamak are presented.

  9. Conceptual design of the tokamak radiation shielding for the Tokamak Physics Experiment (TPX)

    SciTech Connect

    Cole, M.J.; Nelson, B.E.; Jones, G.H.; Goranson, P.L.; Gohar, Y.; Liew, S.L.

    1993-11-01

    The tokamak radiation shielding includes the neutron and gamma shielding around the torus and penetrations required to (1) limit activation of components outside the shield to levels that permit hands-on maintenance and (2) limit the nuclear heating of the superconducting coils and cold structure. The primary design drivers are space, the 350{degree}C bakeout temperature, and cost; therefore, different shield materials were used for different shield components and locations. The shielding is divided into three areas: (1) torus shielding around the vacuum vessel, (2) duct shielding around the vacuum pumping ducts and vertical diagnostic ducts, and (3) penetration shielding in and around the radial ports. The major shield components include water between the walls of the vacuum vessel, lead monosilicate/boron carbide tiles that are attached to the exterior of the vacuum vessel, shield plugs that rill the openings of the large radial ports, and polyethylene/lead/boron shield blocks for duct shielding. A description of the shielding configuration and the performance and operational requirements will be discussed.

  10. Conceptual design of the tokamak radiation shielding for the Tokamak Physics Experiment (TPX)

    NASA Astrophysics Data System (ADS)

    Cole, M. J.; Nelson, B. E.; Jones, G. H.; Goranson, P. L.; Gohar, Y.; Liew, S. L.

    The tokamak radiation shielding includes the neutron and gamma shielding around the torus and penetrations required to (1) limit activation of components outside the shield to levels that permit hands-on maintenance, and (2) limit the nuclear heating of the superconducting coils and cold structure. The primary design drivers are space, the 350 C bakeout temperature, and cost; therefore, different shield materials were used for different shield components and locations. The shielding is divided into three areas: (1) torus shielding around the vacuum vessel, (2) duct shielding around the vacuum pumping ducts and vertical diagnostic ducts, and (3) penetration shielding in and around the radial ports. The major shield components include water between the walls of the vacuum vessel, lead monosilicate/boron carbide tiles that are attached to the exterior of the vacuum vessel, shield plugs that fill the openings of the large radial ports, and polyethylene/lead/boron shield blocks for duct shielding. A description of the shielding configuration and the performance and operational requirements are discussed.

  11. Recent Progress on Spherical Torus Research

    SciTech Connect

    Ono, Masayuki; Kaita, Robert

    2014-01-01

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

  12. Enhanced confinement regimes and control technology in the DIII-D tokamak

    SciTech Connect

    Lohr, J.; Burrell, K.H.; Coda, S.

    1993-07-01

    Advanced tokamak performance has been demonstrated in the DIII-D tokamak in a series of experiments which brought together developments in technology and improved understanding of the physical principles underlying tokamak operation. The achievement of greatly improved confinement coupled with development of new systems for real time plasma control have permitted investigation of the heretofore hidden or poorly controlled variables which together determine global confinement. These experiments, which included work in transport and control of the plasma boundary, point toward development of operationally and economically attractive reactors based on the tokamak. Some of these experiments are described.

  13. Plasma shaping effects on tokamak scrape-off layer turbulence

    NASA Astrophysics Data System (ADS)

    Riva, Fabio; Lanti, Emmanuel; Jolliet, Sébastien; Ricci, Paolo

    2017-03-01

    The impact of plasma shaping on tokamak scrape-off layer (SOL) turbulence is investigated. The drift-reduced Braginskii equations are written for arbitrary magnetic geometries, and an analytical equilibrium model is used to introduce the dependence of turbulence equations on tokamak inverse aspect ratio (ε ), Shafranov’s shift (Δ), elongation (κ), and triangularity (δ). A linear study of plasma shaping effects on the growth rate of resistive ballooning modes (RBMs) and resistive drift waves (RDWs) reveals that RBMs are strongly stabilized by elongation and negative triangularity, while RDWs are only slightly stabilized in non-circular magnetic geometries. Assuming that the linear instabilities saturate due to nonlinear local flattening of the plasma gradient, the equilibrium gradient pressure length {L}p=-{p}e/{{\

  14. Three-dimensional equilibria in axially symmetric tokamaks

    PubMed Central

    Garabedian, Paul R.

    2006-01-01

    The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of α particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. PMID:17159158

  15. Three-dimensional analysis of tokamaks and stellarators.

    PubMed

    Garabedian, Paul R

    2008-09-16

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project.

  16. Analytical solutions for Tokamak equilibria with reversed toroidal current

    SciTech Connect

    Martins, Caroline G. L.; Roberto, M.; Braga, F. L.; Caldas, I. L.

    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.

  17. Three-dimensional analysis of tokamaks and stellarators

    PubMed Central

    Garabedian, Paul R.

    2008-01-01

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

  18. Development of magnetohydrodynamic modes during sawteeth in tokamak plasmas

    SciTech Connect

    Firpo, M.-C.; Ettoumi, W.; Farengo, R.; Ferrari, H. E.; García-Martínez, P. L.; Lifschitz, A. F.

    2013-07-15

    A dynamical analysis applied to a reduced resistive magnetohydrodynamics model is shown to explain the chronology of the nonlinear destabilization of modes observed in tokamak sawteeth. A special emphasis is put on the nonlinear self-consistent perturbation of the axisymmetric m = n = 0 mode that manifests through the q-profile evolution. For the very low fusion-relevant resistivity values, the q-profile is shown to remain almost unchanged on the early nonlinear timescale within the central tokamak region, which supports a partial reconnection scenario. Within the resistive region, indications for a local flattening or even a local reversed-shear of the q-profile are given. The impact of this ingredient in the occurrence of the sawtooth crash is discussed.

  19. Multipoint Thomson scattering diagnostic for the ETE tokamak

    NASA Astrophysics Data System (ADS)

    Berni, L. A.; Alonso, M. P.; Oliveira, R. M.

    2004-10-01

    To measure the electron temperature and plasma density profiles on the Experimento Tokamak Esférico tokamak a multiplexed Thomson scattering diagnostic was implemented. The diagnostic is based on a 10 J ruby laser and a single five spectral channel filter polychromator. A collection lens with f/6.3 relay the scattered light from 23 spatial points to optical fibers. The fibers have a monotonous increasing length and are inserted into the polychromator. Between the collection lens and each fiber optic we have a microlens to match the numerical aperture and to enlarge the plasma observation volume. This work describes the project, the simulations, and the preliminary results obtained with the first four optical fibers.

  20. Neural network prediction of some classes of tokamak disruptions

    NASA Astrophysics Data System (ADS)

    Hernandez, J. V.; Vannucci, A.; Tajima, T.; Lin, Z.; Horton, W.; McCool, S. C.

    1996-08-01

    The use of neural network algorithms for predicting minor and major disruptions in tokamaks is explored by analysing disruption data from the TEXT tokamak with two network architectures. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal measured by a single Mirnov coil. The time step used (=0.04 ms) corresponds to the experimental data sampling rate. Two kinds of approach are adopted for the network: the contiguous future prediction and the multi-time-scale prediction. Both networks are trained through the back-propagation algorithm with inertial terms and the strengths of the results are compared. The use of additional diamagnetic signals as a method of increasing the performance is suggested. The degree of success indicates that the magnetic fluctuations associated with the TEXT disruption data may be characterized by a low dimensional dynamical system

  1. Tokamak Plasma Flows Induced by Local RF Forces

    NASA Astrophysics Data System (ADS)

    Chen, Jiale; Gao, Zhe

    2015-10-01

    The tokamak plasma flows induced by the local radio frequency (RF) forces in the core region are analyzed. The effective components of local RF forces are composed of the momentum absorption term and the resonant parallel momentum transport term (i.e. the parallel component of the resonant ponderomotive forces). Different momentum balance relations are employed to calculate the plasma flows depending on different assumptions of momentum transport. With the RF fields solved from RF simulation codes, the toroidal and poloidal flows by these forces under the lower hybrid current drive and the mode conversion ion cyclotron resonance heating on EAST-like plasmas are evaluated. supported by National Natural Science Foundation of China (Nos. 11405218, 11325524, 11375235 and 11261140327), in part by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB111002, 2013GB112001 and 2013GB112010), and the Program of Fusion Reactor Physics and Digital Tokamak with the CAS “One-Three-Five” Strategic Planning

  2. Pulse length assessment of compact ignition tokamak designs

    SciTech Connect

    Stotler, D.P.; Pomphrey, N.

    1989-07-01

    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 /approx gt/ 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.

  3. Equilibrium calculations for plasma control in CIT (Compact Ignition Tokamak)

    SciTech Connect

    Strickler, D.J.; Peng, Y-K.M. . Fusion Engineering Design Center); Pomphrey, N.; Jardin, S.C. . Plasma Physics Lab.)

    1990-01-01

    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.

  4. Experimental modelling of eddy currents and deflection for tokamak limiters

    SciTech Connect

    Hua, T.Q.; Knott, M.J.; Turner, L.R.; Wehrle, R.B.

    1986-11-01

    During plasma disruptions in a tokamak fusion reactor, eddy currents are induced in the limiters and other conducting structures surrounding the plasma. Interactions between these currents with the toroidal field causes deflection and stress in the structural components. The structural motion in the strong magnetic field induces additional eddy current opposing the initial eddy current and modifying subsequent structural dynamics. Therefore, the motion and current are coupled and must be solved simultaneously. The coupling between current and deflection in cantilevered beams was investigated experimentally. The beams provide a simple model for the limiter blade of a tokamak fusion reactor. Several test pieces and various magnetic field conditions were employed to study the extend of the coupling effect from weak to strong coupling. Experimental results are compared with analytical predictions.

  5. The effects of the magnetic equilibrium on tokamak edge instabilities

    NASA Astrophysics Data System (ADS)

    Wan, Weigang; Chen, Yang; Parker, Scott

    2013-10-01

    The general magnetic equilibrium of the ``full'' geometry is implemented in the global gyrokinetic turbulence code GEM. A mapping between the experimental coordinate (R , Z) and the simulation coordinate (r , θ) is calculated directly from the EQDSK file generated by the EFIT analysis. Tokamak edge simulations are carried out with general geometry. At the edge, the general magnetic equilibria differ significantly from that parametrized by the Miller equilibrium, and this difference has quantitative effects on the linear instabilities of the kinetic peeling ballooning mode and the kinetic ballooning mode found in our previous studies. The growth rates are quite sensitive to the magnetic equilibrium, especially with the Miller quantities of elongation and triangularity. Additionally, the calculation of the safety factor has great uncertainty near the separatrix, and this uncertainty may have important effects on tokamak stability.

  6. Numerical Study of Tokamak Equilibrium with Toroidal Flow on EAST

    NASA Astrophysics Data System (ADS)

    Ren, Qilong; Zhang, Cheng

    2006-09-01

    The effect of the toroidal flow on the equilibrium of tokamak plasmas is a sensitive point for high performance plasma and its precise control. In this paper the effect is studied numerically using the EFIT (Equilibrium Fitting) code on EAST (Experimental Advanced Superconducting Tokamak). Firstly, the numerical calculation exhibits a clear outward shift of pressure contour from the magnetic surfaces in the plasma core and the shift grows with the increase of the toroidal velocity. The peak shift of 8% is observed when the ratio between the plasma velocity and the Alfvén speed equals to 0.15. Secondly, it is shown that the magnetic surfaces shift outwards from those without flow. With a certain plasma current the safety factor on the magnetic axis decreases as the plasma flow velocity increases. The magnetic shear increases about 10% on the plasma boundary compared with the case without flow.

  7. Effects of Magnetic Measurement Uncertainty on Tokamak Equilibrium Reconstruction

    NASA Astrophysics Data System (ADS)

    Montgomery, A. L.; Lao, L. L.; Strait, E. J.; Taylor, T. S.

    2006-10-01

    Reconstruction of the magnetic topology and current density profile from external magnetic diagnostics is vital to the study of tokamak plasmas. It has long been suspected that external magnetic measurements may contain information about the current density near the edge of a shaped plasma. To better reconstruct this feature, the sensitivity of the reconstruction to uncertainty in magnetic measurements must be determined. This is done by analyzing existing DIII-D data for the effect of measurement uncertainty on the edge current density and the location of the separatrix. The new magnetic uncertainty matrix recently added to EFIT provides a basis for this numerical analysis, and these calculations are compared with simple analytical models. This study begins with examination of measurement error in the tokamak with no plasma, and continues to the more complicated plasma scenario. The relative importance of the various measurements can be determined, with the divertor X-point likely to be the most sensitive to the edge current density.

  8. Equilibrium analysis of tokamak discharges with toroidal variation

    SciTech Connect

    Zwingmann, W.; Becoulet, M.; Moreau, Ph.; Nardon, E.

    2006-11-30

    Tokamaks provide a field structure that is almost axisymmetric around the torus axis. There are however always small toroidal variations due to the limited number of toroidal field coils, the magnetic field ripple. On the other hand, non-axisymmetric external fields are applied on purpose to ergodise the field structure close to the separatrix, to control the heat and particle transport across the plasma boundary. We present a perturbation method to calculate the magnetic field of tokamak discharges with with weak toroidal variation. The method is applied for the equilibrium reconstruction of Tore Supra discharges with toroidal ripple. The perturbation method does not rely on a flux surface representation and can therefore be applied to structures with magnetic islands. We obtain the plasma response to the field of ergodising external coils, as proposed for the ITER device.

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

    SciTech Connect

    Zimin, A. M.; Krupin, V. A.; Troynov, V. I.; Klyuchnikov, L. A.

    2015-12-15

    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 (d{sup 3}Π{sub u}–a{sup 3}Σ{sub g}{sup +}) of deuterium in the wavelength range from 590 to 640 nm. The rotational temperatures of ground state X{sup 1}Σ{sub g}{sup +} and upper excited state d{sup 3}Π{sub u} are estimated by the measured spectra.

  10. Maintenance concept development for the Compact Ignition Tokamak

    SciTech Connect

    Macdonald, D.

    1988-01-01

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

  11. Collisionless microtearing modes in hot tokamaks: Effect of trapped electrons

    SciTech Connect

    Swamy, Aditya K.; Ganesh, R.; Brunner, S.; Vaclavik, J.; Villard, L.

    2015-07-15

    Collisionless microtearing modes have recently been found linearly unstable in sharp temperature gradient regions of large aspect ratio tokamaks. The magnetic drift resonance of passing electrons has been found to be sufficient to destabilise these modes above a threshold plasma β. A global gyrokinetic study, including both passing electrons as well as trapped electrons, shows that the non-adiabatic contribution of the trapped electrons provides a resonant destabilization, especially at large toroidal mode numbers, for a given aspect ratio. The global 2D mode structures show important changes to the destabilising electrostatic potential. The β threshold for the onset of the instability is found to be generally downshifted by the inclusion of trapped electrons. A scan in the aspect ratio of the tokamak configuration, from medium to large but finite values, clearly indicates a significant destabilizing contribution from trapped electrons at small aspect ratio, with a diminishing role at larger aspect ratios.

  12. Long slide-away discharges in the COMPASS tokamak

    NASA Astrophysics Data System (ADS)

    Ficker, Ondrej; Mlynar, Jan; Vlainic, Milos; Weinzettl, Vladimir; Urban, Jakub; Cavalier, Jordan; Havlicek, Jaroslav; Panek, Radomir; Hron, Martin; Cerovsky, Jaroslav; Vondracek, Petr; Paprok, Richard; Decker, Joan; Peysson, Yves; Bogar, Ondrej; Stahl, Adam; Compass Team

    2016-10-01

    In this contribution, long runaway electron (RE) dominated discharges achieved in the COMPASS tokamak are presented. The extensive length is possible due to a low consumption of available volt-seconds of the tokamak transformer in this type of discharge. Energetic electron losses in this regime seems to be modulated mainly by small oscillations of a radial position (controller setting) unlike in the RE discharges at higher electron density, where various MHD phenomena affect the evolution of the losses. The behaviour of the slide-away plasma is studied using magnetic coils, HXR detectors, ECE system and a pair of 3He proportional counters of neutrons. The plasma scenario is also modelled using Fokker-Planck codes. EUROfusion WP MST1, MST2.

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

    SciTech Connect

    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.

  14. Residual zonal flows in tokamaks and stellarators at arbitrary wavelengths

    NASA Astrophysics Data System (ADS)

    Monreal, Pedro; Calvo, Iván; Sánchez, Edilberto; Parra, Félix I.; Bustos, Andrés; Könies, Axel; Kleiber, Ralf; Görler, Tobias

    2016-04-01

    In the linear collisionless limit, a zonal potential perturbation in a toroidal plasma relaxes, in general, to a non-zero residual value. Expressions for the residual value in tokamak and stellarator geometries, and for arbitrary wavelengths, are derived. These expressions involve averages over the lowest order particle trajectories, that typically cannot be evaluated analytically. In this work, an efficient numerical method for the evaluation of such expressions is reported. It is shown that this method is faster than direct gyrokinetic simulations performed with the Gene and EUTERPE codes. Calculations of the residual value in stellarators are provided for much shorter wavelengths than previously available in the literature. Electrons must be treated kinetically in stellarators because, unlike in tokamaks, kinetic electrons modify the residual value even at long wavelengths. This effect, that had already been predicted theoretically, is confirmed by gyrokinetic simulations.

  15. [alpha]-particle transport-driven current in tokamaks

    SciTech Connect

    Heikkinen, J.A. ); Sipilae, S.K. )

    1995-03-01

    It is shown that the radial transport of fusion-born energetic [alpha] particles, induced by electrostatic waves traveling in one poloidal direction, is directly connected to a net momentum of [alpha] particles in the toroidal direction in tokamaks. Because the momentum change is almost independent of toroidal velocity, the energy required for the momentum generation remains small on an [alpha]-particle population sustained by an isotropic time-independent source. By numerical toroidal Monte Carlo calculations it is shown that the current carried by [alpha] particles in the presence of intense well penetrated waves can reach several mega-amperes in reactor-sized tokamaks. The current obtained can greatly exceed the neoclassical bootstrap current of the [alpha] particles.

  16. Theory of magnetohydrodynamic instabilities excited by energetic particles in tokamaks

    SciTech Connect

    Chen, L. )

    1994-07-20

    The resonant excitations of high-n magnetohydrodynamic (MHD) instabilities by the energetic ions/alpha particles in tokamaks are theoretically analyzed. Here, n is the toroidal mode number. Since, typically, the MHD modes consist of two-scale structures; one singular ( inertial'') region and one regular (ideal) region, the energetic particle contributions in the singular region are suppressed by the finite-size orbits. Analytical dispersion relations can then be derived via the asymptotic matching analysis. The dispersion relations have the generic form of the fishbone'' dispersion relation and demonstrate, in particular, the existence of two types of modes; that is, the MHD gap mode and the energetic-particle continuum mode. Specific expressions are given for both the kinetic ballooning modes (KBM) and the toroidal Alfven modes (TAM). It is suggested that the stability property may be qualitatively regarded as the hybrid'' of conventional MHD tokamaks and field reversed ion rings. [copyright]American Institute of Physics

  17. The Aneutronic Rodless Ultra Low Aspect Ratio Tokamak

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2016-10-01

    The replacement of the metal centre-post in spherical tokamaks (STs) by a plasma centre-post (PCP, the TF current carrier) is the ideal scenario for a ST reactor. A simple rodless ultra low aspect-ratio tokamak (RULART) using a screw-pinch PCP ECR-assisted with an external solenoid has been proposed in the most compact RULART [Ribeiro C, SOFE-15]. There the solenoid provided the stabilizing field for the PCP and the toroidal electrical field for the tokamak start-up, which will stabilize further the PCP, acting as stabilizing closed conducting surface. Relative low TF will be required. The compactness (high ratio of plasma-spherical vessel volume) may provide passive stabilization and easier access to L-H mode transition. It is presented here: 1) stability analysis of the PCP (initially MHD stable due to the hollow J profile); 2) tokamak equilibrium simulations, and 3) potential use for aneutronic reactions studies via pairs of proton p and boron 11B ion beams in He plasmas. The beams' line-of-sights sufficiently miss the sources of each other, thus allowing a near maximum relative velocities and reactivity. The reactions should occur close to the PCP mid-plane. Some born alphas should cross the PCP and be dragged by the ion flow (higher momentum exchange) towards the anode but escape directly to a direct electricity converter. Others will reach evenly the vessel directly or via thermal diffusion (favourable heating by the large excursion 2a), leading to the lowest power wall load possible. This might be a potential hybrid direct-steam cycle conversion reactor scheme, nearly aneutronic, and with no ash or particle retention problems, as opposed to the D-T thermal reaction proposals.

  18. Thermally excited proton spin-flip laser emission in tokamaks

    SciTech Connect

    Arunasalam, V.; Greene, G.J.

    1993-07-01

    Based on statistical thermodynamic fluctuation arguments, it is shown here for the first time that thermally excited spin-flip laser emission from the fusion product protons can occur in large tokamak devices that are entering the reactor regime of operation. Existing experimental data from TFTR supports this conjecture, in the sense that these measurements are in complete agreement with the predictions of the quasilinear theory of the spin-flip laser.

  19. Momentum injection in tokamak plasmas and transitions to reduced transport.

    PubMed

    Parra, F I; Barnes, M; Highcock, E G; Schekochihin, A A; Cowley, S C

    2011-03-18

    The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.

  20. Viscous damping of toroidal angular momentum in tokamaks

    SciTech Connect

    Stacey, W. M.

    2014-09-15

    The Braginskii viscous stress tensor formalism was generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry in order to provide a representation for the viscous damping of toroidal rotation in tokamaks arising from various “neoclassical toroidal viscosity” mechanisms. In the process, it was verified that the parallel viscosity contribution to damping toroidal angular momentum still vanishes even in the presence of toroidal asymmetries, unless there are 3D radial magnetic fields.

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

    NASA Astrophysics Data System (ADS)

    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.

  2. Tokamak Startup Using Point-Source dc Helicity Injection

    SciTech Connect

    Battaglia, D. J.; Bongard, M. W.; Fonck, R. J.; Redd, A. J.; Sontag, A. C.

    2009-06-05

    Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.

  3. Tokamak magnetic islands in the presence of nonaxisymmetric perturbations

    SciTech Connect

    Reiman, A.H.

    1991-07-01

    The effects of a small, externally imposed, nonaxisymmetric magnetic field perturbation on magnetic islands are studied analytically, assuming zero {beta}, tokamak ordering, and narrow islands. For the tearing stable case, the conditions under which the self-consistent plasma response is self-healing or amplifying are elucidated. For the tearing unstable case, the quasilinear theory of tearing modes is extended to a description of locked modes. 16 refs., 12 figs.

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

    NASA Astrophysics Data System (ADS)

    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.

  5. A moving finite element model of the tokamak scrapeoff layer

    SciTech Connect

    Glasser, A.H.; Kuprat, A.P.

    1993-10-01

    Most numerical simulations of the tokamak scrapeoff layer use a mapping to flux coordinates and a piecewise equidistributed grid in those coordinates to resolve the multiple length scales and anisotropy characteristic of this problem. We have developed an alternative numerical method using simple cylindrical coordinates with a complex adaptive grid scheme. It is based on an understructured grid of traingles which move adaptively, aligning themselves with the magnetic field and concentrating in regions of sharp gradients.

  6. Flux generation in ultra-low- q tokamak discharges

    SciTech Connect

    Schmid, P.; Barrick, G.; Greene, P. ); Robertson, S. )

    1991-04-01

    Spontaneous generation of toroidal flux is observed in a reversed-field pinch device operated as an ultra-low-{ital q} tokamak with a safety factor on axis of 1/4. An initial toroidal bias field of 150 G is increased to 600 G on axis in 300 {mu}sec while the field at the wall is held nearly constant. Sawteeth are observed, which indicate cyclic magnetic reconnection.

  7. Momentum Injection in Tokamak Plasmas and Transitions to Reduced Transport

    SciTech Connect

    Parra, F. I.; Highcock, E. G.; Schekochihin, A. A.; Barnes, M.

    2011-03-18

    The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.

  8. Design of geometric phase measurement in EAST Tokamak

    NASA Astrophysics Data System (ADS)

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

    2016-07-01

    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.

  9. Economic analyses of alpha channeling in tokamak power plants.

    SciTech Connect

    Ehst, D.A.

    1998-09-17

    The hot-ion-mode of operation [1] has long been thought to offer optimized performance for long-pulse or steady-state magnetic fusion power plants. This concept was revived in recent years when theoretical considerations suggested that nonthermal fusion alpha particles could be made to channel their power density preferentially to the fuel ions [2,3]. This so-called anomalous alpha particle slowing down can create plasmas with fuel ion temperate T{sub i} somewhat larger than the electron temperature T{sub e}, which puts more of the beta-limited plasma pressure into the useful fuel species (rather than non-reacting electrons). As we show here, this perceived benefit may be negligible or nonexistent for tokamaks with steady state current drive. It has likewise been argued [2,3] that alpha channeling could be arranged such that little or no external power would be needed to generate the steady state toroidal current. Under optimistic assumptions we show that such alpha-channeling current drive would moderately improve the economic performance of a first stability tokamak like ARIES-I [4], however a reversed-shear (advanced equilibrium) tokamak would likely not benefit since traditional radio-wave (rf) electron-heating current drive power would already be quite small.

  10. Simulation of MST tokamak discharges with resonant magnetic perturbations

    NASA Astrophysics Data System (ADS)

    Cornille, B. S.; Sovinec, C. R.; Chapman, B. E.; Dubois, A.; McCollam, K. J.; Munaretto, S.

    2016-10-01

    Nonlinear MHD modeling of MST tokamak plasmas with an applied resonant magnetic perturbation (RMP) reveals degradation of flux surfaces that may account for the experimentally observed suppression of runaway electrons with the RMP. Runaway electrons are routinely generated in MST tokamak discharges with low plasma density. When an m = 3 RMP is applied these electrons are strongly suppressed, while an m = 1 RMP of comparable amplitude has little effect. The computations are performed using the NIMROD code and use reconstructed equilibrium states of MST tokamak plasmas with q (0) < 1 and q (a) = 2.2 . Linear computations show that the (1 , 1) -kink and (2 , 2) -tearing modes are unstable, and nonlinear simulations produce sawtoothing with a period of approximately 0.5 ms, which is comparable to the period of MHD activity observed experimentally. Adding an m = 3 RMP in the computation degrades flux surfaces in the outer region of the plasma, while no degradation occurs with an m = 1 RMP. The outer flux surface degradation with the m = 3 RMP, combined with the sawtooth-induced distortion of flux surfaces in the core, may account for the observed suppression of runaway electrons. Work supported by DOE Grant DE-FC02-08ER54975.

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

    SciTech Connect

    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. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    SciTech Connect

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    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 Nb/sub 3/Sn 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.

  13. Transition to subcritical turbulence in a tokamak plasma

    NASA Astrophysics Data System (ADS)

    van Wyk, F.; Highcock, E. G.; Schekochihin, A. A.; Roach, C. M.; Field, A. R.; Dorland, W.

    2016-12-01

    Tokamak turbulence, driven by the ion-temperature gradient and occurring in the presence of flow shear, is investigated by means of local, ion-scale, electrostatic gyrokinetic simulations (with both kinetic ions and electrons) of the conditions in the outer core of the Mega-Ampere Spherical Tokamak (MAST). A parameter scan in the local values of the ion-temperature gradient and flow shear is performed. It is demonstrated that the experimentally observed state is near the stability threshold and that this stability threshold is nonlinear: sheared turbulence is subcritical, i.e. the system is formally stable to small perturbations, but, given a large enough initial perturbation, it transitions to a turbulent state. A scenario for such a transition is proposed and supported by numerical results: close to threshold, the nonlinear saturated state and the associated anomalous heat transport are dominated by long-lived coherent structures, which drift across the domain, have finite amplitudes, but are not volume filling; as the system is taken away from the threshold into the more unstable regime, the number of these structures increases until they overlap and a more conventional chaotic state emerges. Whereas this appears to represent a new scenario for transition to turbulence in tokamak plasmas, it is reminiscent of the behaviour of other subcritically turbulent systems, e.g. pipe flows and Keplerian magnetorotational accretion flows.

  14. 3D Equilibrium Reconstruction in Stellarators and Tokamaks with STELLOPT

    NASA Astrophysics Data System (ADS)

    Lazerson, Samuel; Pablant, Novimir; Gates, David; Neilson, Hutch; Nazikian, Raffi; Suzuki, Yasuhiro; Watanabe, Kiyomasa; Ida, Katsumi; Sakakibara, Satoru

    2012-10-01

    The ability to model and predict the behavior of stellarators and tokamaks requires an ability to match simulation parameters with experimental measurements. This process, known as experimental reconstruction, has been used extensively with 2D axisymmetric codes for Tokamaks. These codes, such as EFIT, lack the ability to model the 3D nature of stellarators and the emerging 3D nature of Tokamaks. Phenomena such as, shielding of islands by neoclassical flows and the suppression of edge localized modes through application of 3D fields, highlight the need for such 3D tools. The stellarator optimizer code STELLOPT has been modified to match 3D VMEC equilibria to experimental measurements. This has allowed 3D experimental reconstructions to be preformed on W7-AS, LHD, and DIII-D devices. The free boundary VMEC equilibria are matched to Thomson profiles (ne and Te), charge exchange measurements (Ti), MSE (polarization angle), and magnetic diagnostics (B-probes, flux loops, Rogowski coils). Three dimensional reconstructed equilibria are presented alongside confidence metrics for the reconstruction process.

  15. The D3-D tokamak trouble report database

    NASA Astrophysics Data System (ADS)

    Petersen, P. I.; Miller, S. M.

    1991-11-01

    Operation of the DIII-D tokamak at General Atomics involves many groups which work on the various subsystems. To overview and speed the solution to trouble or problem areas that limit machine availability, a common trouble report system was established. The TROUBLE database automates the recording of trouble reports and eases analysis of problem areas. It contains information on equipment affected, description of problem, cause of problem, solution to problem, and machine downtime (if any). It was created using S1032 from Compuserve Data Technologies and runs on a VAX 8650. The data is used to find the major problem areas so they can be solved and improve the tokamak availability. The data is available to Idaho National Engineering Laboratory (INEL). They are using the data with data from other tokamaks to develop a Fusion Failure Experience Data Collection. Our experience is that a few failures are often the cause of a major part of the downtime. We will discuss these failures and the actions taken to correct them. The data base also will be used to determine the preventive maintenance schedule for different components.

  16. Shape reconstruction of merging spherical tokamak plasma in UTST device

    NASA Astrophysics Data System (ADS)

    Ushiki, Tomohiko; Itagaki, Masafumi; Inomoto, Michiaki

    2016-10-01

    Spherical tokamak (ST) merging method is one of the ST start-up methods which heats the plasma through magnetic reconnection. In the present study reconstruction of eddy current profile and plasma shape was performed during spherical tokamak merging only using external sensor signals by the Cauchy condition surface (CCS) method. CCS method have been implemented for JT-60 (QST), QUEST (Kyushu University), KSTAR (NFRI), RELAX (KIT), and LHD (Nifs). In this method, CCS was assumed inside each plasmas, where both flux function and its normal derivative are unknown. Effect of plasma current was replaced by the boundary condition of CCS, assuming vacuum field everywhere. Also, the nodal points for the boundary integrals of eddy current density were set using quadratic elements in order to express the complicated vacuum vessel shape. Reconstructed profiles of the eddy current and the magnetic flux were well coincided with the reference in each phase of merging process. Magnetic sensor installation plan for UTST was determined from these calculation results. This work was supported by the JSPS A3 Foresight Program ``Innovative Tokamak Plasma Startup and Current Drive in Spherical Torus''.

  17. Observation of Pedestal Plasma Turbulence on EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Gao, Xiang; Zhang, Tao; Han, Xiang; Zhang, Shoubiao; Wang, Yumin; Liu, Zixi; Yang, Yao; Liu, Shaocheng; Shi, Nan; Ling, Bili; Li, Jiangang; The EAST Team

    2013-08-01

    Pedestal plasma turbulence was experimentally studied by microwave reflectometry on EAST tokamak. The characteristics of edge pedestal turbulence during dithering L-H transition, ELM-free H-mode phase and inter-ELM phase have recently been studied on EAST. An edge spatial structure of density fluctuation and its dithering temporal evolution is observed for the first time on the EAST tokamak during the L-H transition phase. A coherent mode usually appears during the ELM-free phase prior to the first ELM on EAST tokamak. The mode frequency gradually decreases as the pedestal evolves. Analysis shows that the coherent mode is in the pedestal region inside the separatrix. In plasma with type-III ELMs, a precursor mode before ELM is usually observed. The frequency of the precursor was initially about 150 kHz and gradually decreased till the next ELM. The mode amplitude increases or shows saturation before ELM. In the plasma with compound ELMs composed of high and low frequency ELMs, the precursor was also observed before the high frequency ELM while the harmonic oscillations with frequencies of 20 kHz, 40 kHz and 60 kHz appear before the low frequency ELM.

  18. Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

    NASA Astrophysics Data System (ADS)

    Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

    2013-10-01

    The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

  19. Microwave Imaging Reflectometer (MIR) Development for the EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Domier, Calvin; Hu, Xing; Spear, Alexander; Zhu, Yilun; Xie, Jinlin; Luhmann, Neville

    2016-10-01

    An upgraded MIR system is being developed for the EAST tokamak based on the successful DIII-D MIR system. The EAST MIR system has 8 radial channels consisting of 8 independent probing frequencies ranging from 75 to 103 GHz, driven by fast tuning synthesizers and active frequency multipliers. There are 12 poloidal channels in the heterodyne down-conversion receiver system, with each channel corresponding to a separate poloidal position inside the tokamak. The down-conversion electronics are designed to optimize signal to noise ratio and are embedded with a microcontroller to realize remote computer control. Considerable improvements are also seen in the front-end plasma facing optics. This new optical system provides features including focusing, zoom, field curvature adjustment, and incident angle adjustment. These functions can be realized together or independently depending on the configuration setup of the large aperture lenses. This MIR system is expected to be installed on the EAST tokamak in December 2016, co-located with the Electron Cyclotron Emission Imaging (ECEI) system, to simultaneously measure electron density and temperature fluctuations. This work was supported by U.S. DOE Grant DE-FG02-99ER54531 and by the National MCF energy development program of China.

  20. Dust-Particle Transport in Tokamak Edge Plasmas

    SciTech Connect

    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.

  1. Experimental verification of subcooled flow boiling for tokamak pump limiter designs

    SciTech Connect

    Koski, J.A.; Beattie, A.G.; Whitley, J.B.; Croessmann, C.D.

    1987-01-01

    In fusion energy research devices such as tokamaks, limiters are used to define the plasma boundary, and may serve the additional functions of plasma density and impurity control by removing neutralized particles from the plasma edge region. Because the devices must operate in the plasma edge, they are subject to high heat fluxes. In this paper, experimental studies conducted in support of a pump limiter design currently under development are discussed. Subcooled flow boiling of water and twisted tape flow enhancement are combined to enable heat removal of highly peaked local heat fluxes at the tube-water boundary in the 40 to 50 MW/m/sup 2/ range. Critical heat flux and heat removal experiments were conducted on copper tube targets with the use of a rastered 30 kV electron beam apparatus capable of producing the desired steady state heat flux levels.

  2. Data acquisition and processing system of the electron cyclotron emission imaging system of the KSTAR tokamak.

    PubMed

    Kim, J B; Lee, W; Yun, G S; Park, H K; Domier, C W; Luhmann, N C

    2010-10-01

    A new innovative electron cyclotron emission imaging (ECEI) diagnostic system for the Korean Superconducting Tokamak Advanced Research (KSTAR) produces a large amount of data. The design of the data acquisition and processing system of the ECEI diagnostic system should consider covering the large data production and flow. The system design is based on the layered structure scalable to the future extension to accommodate increasing data demands. Software architecture that allows a web-based monitoring of the operation status, remote experiment, and data analysis is discussed. The operating software will help machine operators and users validate the acquired data promptly, prepare next discharge, and enhance the experiment performance and data analysis in a distributed environment.

  3. Steady state off-axis sawtoothing in the Rijnhuizen Tokamak project

    NASA Astrophysics Data System (ADS)

    Meulenbroeks, R. F. G.; de Baar, M. R.; Beurskens, M. N. A.; de Blank, H. J.; Deng, B. H.; Donné, A. J. H.; Hogeweij, G. M. D.; Lopes Cardozo, N. J.; Montvai, A.; Oyevaar, Th.

    1999-10-01

    A family of off-axis, or annular, instabilities has been studied using Thomson scattering, soft X-ray emission, and two electron cyclotron emission diagnostic systems. In the Rijnhuizen tokamak (RTP) [N. J. Lopes Cardozo et al., Plasma Physics and Controlled Nuclear Fusion Research 1992 (International Atomic Energy Agency, Vienna, 1993), Vol. 1, p. 271] these phenomena are invoked in a controlled way in discharges with specific (off-axis) deposition of electron cyclotron heating (ECH) and persist during most of the heating period, or during many current diffusion times. Based on coherent mode analysis at the crash time, the instabilities are associated with resonant surfaces near simple rational values of q (3/2, 2, and 3). A parameter study shows an increase of reheat rate and a decrease of sawtooth period with increasing ECH power and — in contrast to observations in other experiments — with increASING density as well.

  4. Sawtooth stabilization by localized electron cyclotron heating in the WT-3 tokamak

    SciTech Connect

    Hanada, K.; Maehara, T.; Makino, K.; Kishigami, Y.; Kishino, T.; Minami, T.; Tanaka, H.; Iida, M.; Nakamura, M.; Maekawa, T.; Terumichi, Y.; Tanaka, S. )

    1992-11-01

    The effect on sawtooth oscillations (STO) by localized electron-cyclotron-resonance heating (ECH) on the WT-3 tokamak ({ital Plasma} {ital Physics} {ital and} {ital Controlled} {ital Nuclear} {ital Fusion} {ital Research}, 1988 (International Atomic Energy Agency, Vienna, 1989), Vol. 1, p. 563) is studied. STO are strongly modified or stabilized by ECH near the {ital q}=1 surface, where {ital q} refers to the safety factor. The effect of ECH is much stronger when it is applied on the high-field side as compared to the low-field side. Further, even when ECH is applied outside the {ital q}=1 surface, the amplitude of STO decreases and STO stabilizes. In the very high {ital q}{sub {ital L}} discharge, the excitation of STO can be obtained by applying ECH.

  5. Application of an array processor to the analysis of magnetic data for the Doublet III tokamak

    SciTech Connect

    Wang, T.S.; Saito, M.T.

    1980-08-01

    Discussed herein is a fast computational technique employing the Floating Point Systems AP-190L array processor to analyze magnetic data for the Doublet III tokamak, a fusion research device. Interpretation of the experimental data requires the repeated solution of a free-boundary nonlinear partial differential equation, which describes the magnetohydrodynamic (MHD) equilibrium of the plasma. For this particular application, we have found that the array processor is only 1.4 and 3.5 times slower than the CDC-7600 and CRAY computers, respectively. The overhead on the host DEC-10 computer was kept to a minimum by chaining the complete Poisson solver and free-boundary algorithm into one single-load module using the vector function chainer (VFC). A simple time-sharing scheme for using the MHD code is also discussed.

  6. High power heating of magnetic reconnection in merging tokamak experiments

    SciTech Connect

    Ono, Y.; Tanabe, H.; Gi, K.; Watanabe, T.; Ii, T.; Yamada, T.; Gryaznevich, M.; Scannell, R.; Conway, N.; Crowley, B.; Michael, C.

    2015-05-15

    Significant ion/electron heating of magnetic reconnection up to 1.2 keV was documented in two spherical tokamak plasma merging experiment on MAST with the significantly large Reynolds number R∼10{sup 5}. Measured 1D/2D contours of ion and electron temperatures reveal clearly energy-conversion mechanisms of magnetic reconnection: huge outflow heating of ions in the downstream and localized heating of electrons at the X-point. Ions are accelerated up to the order of poloidal Alfven speed in the reconnection outflow region and are thermalized by fast shock-like density pileups formed in the downstreams, in agreement with recent solar satellite observations and PIC simulation results. The magnetic reconnection efficiently converts the reconnecting (poloidal) magnetic energy mostly into ion thermal energy through the outflow, causing the reconnection heating energy proportional to square of the reconnecting (poloidal) magnetic field B{sub rec}{sup 2}  ∼  B{sub p}{sup 2}. The guide toroidal field B{sub t} does not affect the bulk heating of ions and electrons, probably because the reconnection/outflow speeds are determined mostly by the external driven inflow by the help of another fast reconnection mechanism: intermittent sheet ejection. The localized electron heating at the X-point increases sharply with the guide toroidal field B{sub t}, probably because the toroidal field increases electron confinement and acceleration length along the X-line. 2D measurements of magnetic field and temperatures in the TS-3 tokamak merging experiment also reveal the detailed reconnection heating mechanisms mentioned above. The high-power heating of tokamak merging is useful not only for laboratory study of reconnection but also for economical startup and heating of tokamak plasmas. The MAST/TS-3 tokamak merging with B{sub p} > 0.4 T will enables us to heat the plasma to the alpha heating regime: T{sub i} > 5 keV without using any additional heating facility.

  7. ADX: a high field, high power density, advanced divertor and RF tokamak

    NASA Astrophysics Data System (ADS)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  8. Millimeter wave tokamak heating and current drive with a high power free electron laser

    SciTech Connect

    Thomassen, K.I.

    1987-01-01

    Experiments on microwave generation using a free electron laser (FEL) have shown this to be an efficient way to generate millimeter wave power in short, intense pulses. Short pulse FEL's have several advantages that make them attractive for application to ECR heating of tokamak fusion reactors. This paper reports on plans made to demonstrate the technology at the Microwave Tokamak Experiment (MTX) Facility.

  9. Transport simulations of ohmic pellet experiments on the TFTR, ASDEX, and ALCATOR-C tokamaks

    SciTech Connect

    Redi, M.H.; Tang, W.M.; Owens, D.K.; Greenwald, M.; Gruber, O.; Kaufmann, M.

    1988-07-01

    Transport simulations of ohmic gas-fuelled and pellet-fuelled experiments have been carried out to test a microinstability-based, profile-consistent model of anomalous transport in tokamaks. Predictions for experiments on the TFTR, ASDEX, and ALCATOR-C tokamaks were found consistent with the observed confinement and temperature measurements. 26 refs., 11 figs., 10 tabs.

  10. The Discharge Design of HL-2M with the Tokamak Simulation Code (TSC)

    SciTech Connect

    Yudong Pan, S.C. Jardin, and C. Kes

    2007-10-10

    We present results on the discharge design of the HL-2M tokamak, which is to be an upgrade to the existing HL-2A tokamak. We present simulation results for complete 5-sec. discharges, both double null and lower single null, for both ohmic and auxiliary heated discharges. We also discuss the vertical stability properties of the device. __________________________________________________

  11. Heavy Neutral Beam Probe for edge plasma analysis in Tokamaks. Annual progress report, December 1, 1992--November 30, 1993

    SciTech Connect

    Castracane, J.; Saravia, E.; Beckstead, J.; Aceto, S.

    1993-09-03

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

  12. A New Method for Shear Stabilization of Advanced Tokamak Reactors via Mode Converted Ion Bernstein Waves*

    NASA Astrophysics Data System (ADS)

    Sund, Richard; Scharer, John

    2002-11-01

    We examine a new method for generating sheared flows in advanced tokamak D-T reactors with the goal of creating and controlling internal transport barriers. Ion-Bernstein waves (IBWs) have the recognized capacity to create internal transport barriers through sheared plasma flows resulting from ion absorption. Under reactor conditions, the IBW can be generated by mode conversion of a fast magnetosonic wave incident from the high-field side (HFS) on the second harmonic resonance of a minority hydrogen component, with near 100200 MHz) minimizes parasitic absorption and permits the converted IBW to approach the fifth tritium harmonic. It also facilitates compact antennas and feeds, and efficient fast wave launch. Placement of the 5T absorption layer on the HFS is advantageous for shear production. The scheme is applicable to reactors with aspect ratio < 3 such that the conversion and absorption layers are both on the high field side of the magnetic axis. Various factors (adequate separation of the mode conversion layer from the magnetic axis, concentration of the fast wave near the midplane, large machine size, and plasma elongation) minimize poloidal field effects in the conversion zone and permit a slab analysis. We use a 1-D full-wave code to analyze the conversion and absorption. A 2-D ray-tracing code incorporating poloidal magnetic fields is used to follow the IBW for various equilibria. Within this analysis a weak bean shape appears most favorable. This is an attractive scheme for future advanced tokamak reactors. *Research supported by the Univ. of Wisconsin, Madison

  13. ATOMIC PHYSICS PROCESSES IMPORTANT TO THE UNDERSTANDING OF THE SCRAPE-OFF LAYER OF TOKAMAKS

    SciTech Connect

    WEST, W.P.; GOLDSMITH,; B. EVANS,T.E.; OLSON, R.J.

    2002-05-01

    The region between the well-confined plasma and the vessel walls of a magnetic confinement fusion research device, the scrape-off layer (SOL), is typically rich in atomic and molecular physics processes. The most advanced magnetic confinement device, the magnetically diverted tokamak, uses a magnetic separatrix to isolate the confinement zone (closed flux surfaces) from the edge plasma (open field lines). Over most of their length the open field lines run parallel to the separatrix, forming a thin magnetic barrier with the nearby vessel walls. In a poloidally-localized region, the open field lines are directed away from the separatrix and into the divertor, a region spatially separated from the separatrix where intense plasma wall interaction can occur relatively safely. Recent data from several tokamaks indicate that particle transport across the field lines of the SOL can be somewhat faster than previously thought. In these cases, the rate at which particles reach the vessel wall is comparable to the rate to the divertor from parallel transport. The SOL can be thin enough that the recycling neutrals and sputtered impurities from the wall may refuel or contaminate the confinement zone more efficiently than divertor plasma wall interaction. Just inside the SOL is a confinement barrier that produces a sharp pedestal in plasma density and temperature. Understanding neutral transport through the SOL and into the pedestal is key to understanding particle balance and particle and impurity exhaust. The SOL plasma is sufficiently hot and dense to excite and ionize neutrals. Ion and neutral temperatures are high enough that charge exchange between the neutrals and fuel and impurity ions is fast. Excitation of neutrals can be fast enough to lead to nonlinear behavior in charge exchange and ionization processes. In this paper the detailed atomic physics important to the understanding of the neutral transport through the SOL will be discussed.

  14. Adaptive feedback control of wall modes in tokamaks

    NASA Astrophysics Data System (ADS)

    Sun, Zhipeng

    The goal of this study is to stabilize the resistive wall modes (RWM) in tokamaks with adaptive stochastic feedback control. This is the first ever attempt at adaptive stochastic feedback optimal control of RWM in tokamaks. Both adaptive optimal state feedback and adaptive output feedback control have been studied. The adaptive optimal state feedback control design successfully stabilizes a slowly time-evolving RWM in a tokamak in a time scale of 4 times the inverse of the growth rate of the RWM. The stabilized system output for the time-invariant model is twice the system noise level. For the time-varying model, it is several times larger than the time-invariant case. The adaptive stochastic output feedback can also stabilize the slowly time-evolving RWM. It can do this in a time about 3 times that of the inverse of the growth rate of the RWM. The stabilized system output is twice as large as that of the state feedback case. In order to avoid the bottleneck encountered in the various sequential computations with big matrices in the feedback algorithms, neural network control has been proposed. It has been used to implement the adaptive stochastic output feedback control. It can stabilize the RWM instability in a time of 3 times the inverse of the growth rate of the RWM. The stabilized wall modes have the steady state output similar to the output feedback case. The developed algorithms, state feedback, output feedback, neural network control, can be readily applied to other plasma instabilities.

  15. Magneto-hydro-dynamic limits in spherical tokamaks

    NASA Astrophysics Data System (ADS)

    Hender, T. C.; Allfrey, S. J.; Akers, R.; Appel, L. C.; Bevir, M. K.; Buttery, R. J.; Gryaznevich, M.; Jenkins, I.; Kwon, O. J.; McClements, K. G.; Martin, R.; Medvedev, S.; Nightingale, M. P. S.; Ribeiro, C.; Roach, C. M.; Robinson, D. C.; Sharapov, S. E.; Sykes, A.; Villard, L.; Walsh, M. J.

    1999-05-01

    The operational limits observed in spherical tokamaks, notably the small tight aspect ratio tokamak (START) device [A. Sykes et al., Nucl. Fusion 32, 694 (1992)], are consistent with those found in conventional aspect ratio tokamaks. In particular the highest β achieved (˜40%) is consistent with an ideal magneto-hydro-dynamic (MHD) Troyon type limit, the upper limit on density is well described by the Greenwald density (πa2n¯e/Ip˜1) and the normalized current (Ip/aBt) is limited such that q95≳2. Stability calculations indicate scope for increasing both normalized β and normalized current beyond the values so far achieved, although wall stabilization is generally needed for low-n modes. In double null configurations current terminating disruptions occur at each of the operational boundaries, though the current quench tends to be slow at the density limit and disruptions at high β may be due to the low q. In early limiter START discharges, before the divertor coils were installed, disruptions rarely occurred. Instead internal reconnection events which have all the characteristics of a disruption except the current quench occurred. These various disruptive behaviors are explained in terms of a model in which helicity is conserved during the disruption. Due to the low toroidal field beam ions in START, and α particles in a ST power plant, are super-Alfvénic. This gives the possibility for toroidal Alfvén eigenmodes (TAEs) to occur and such modes are frequently observed in START neutral beam injection (NBI) discharges, but seem to be benign. The features of these observed TAEs are shown to be in agreement with MHD calculations.

  16. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    SciTech Connect

    Simonen, T.C.; Baker, D.

    1993-01-01

    The DIII-D tokamak research program is carried out by General Atomics for the U.S. Department of Energy. The DIII-D is the most flexible and best diagnosed tokamak in the world and the second largest tokamak in the U.S. The primary goal of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated in three major areas: Tokamak Physics, Divertor and Boundary Physics, and Advanced Tokamak Studies.

  17. DIII-D research operations. Annual report, October 1, 1992--September 30, 1993

    SciTech Connect

    La Haye, R.J.

    1994-05-01

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

  18. Nonlinear tearing instabilities in tokamaks with locally flattened current profiles

    SciTech Connect

    Reiman, A.H.

    1988-07-01

    Nonlinear tearing stability is evaluated for current profiles which are linearly stabilized by flattening the current in the neighborhood of the rational surface. When marginally stable to the linear instability, these profiles remain unstable in the presence of a small but finite island. The growth of the island saturated only when the island reaches the width it would have attained in the absence of flattening. Implications are discused for proposed methods of tearing mode stabilization and for theories of the tokamak sawtooth oscillation. 19 refs., 1 fig.

  19. Poloidal flow damping with potato orbits in tokamaks

    SciTech Connect

    Shaing, K.C.

    2005-10-01

    The poloidal flow damping rate in the vicinity of the magnetic axis in tokamaks is calculated using the time-dependent plasma viscosity. It is found that the damping rate is of the order of {nu}{sub ii}/f{sub t}{sup 2}, where {nu}{sub ii} is the ion-ion collision frequency, and f{sub t} is the fraction of the trapped potatoes. The corresponding neoclassical polarization or inertia enhancement factor is [1+({sigma}{sub p}q{sup 2}/f{sub t})], where {sigma}{sub p} is a numerical number of the order of unity, and q is the safety factor.

  20. Ion plateau transport near the tokamak magnetic axis

    SciTech Connect

    Shaing, K.C.; Hazeltine, R.D.

    1998-04-01

    Conventional neoclassical transport theory does not pertain near the magnetic axis, where orbital variation of the minor radius and the poloidal field markedly change the nature of guiding-center trajectories. Instead of the conventional tokamak banana-shaped trajectories, near-axis orbits, called potato orbits, are radially wider and lead to distinctive kinetic considerations. Here it is shown that there is a plateau regime for the near-axis case; the corresponding potato-plateau ion thermal conductivity is computed. {copyright} {ital 1998 American Institute of Physics.}

  1. Resistive toroidal-field coils for tokamak reactors

    SciTech Connect

    Kalnavarns, J.; Jassby, D.L.

    1980-11-01

    This paper analyzes the optimization of the geometry of resistive TF coils of rectangular bore for tokamak fusion test reactors and practical neutron generators. In examining the trade-offs between geometric parameters and magnetic field for reactors giving a specified neutron wall loading, either the resistive power loss or the lifetime coil cost can be minimized. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs. Bending moment distributions in closed form have been derived for rectangular coils on the basis of the theory of rigid frames. Candidate methods of fabrication and of implementing demountable joints are summarized.

  2. Tokamak transmutation of (nuclear) waste (TTW): Parametric studies

    SciTech Connect

    Cheng, E.T.; Krakowski, R.A.; Peng, Y.K.M.

    1994-06-01

    Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low-aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses.

  3. Sensitivity of kinetic ballooning mode instability to tokamak equilibrium implementations

    NASA Astrophysics Data System (ADS)

    Xie, H. S.; Xiao, Y.; Holod, I.; Lin, Z.; Belli, E. A.

    2016-10-01

    Global, first-principles study of the kinetic ballooning mode (KBM) is crucial to understand tokamak edge physics in high-confinement mode (H-mode). In contrast to the ion temperature gradient mode and trapped electron mode, the KBM is found to be very sensitive to the equilibrium implementations in gyrokinetic codes. In this paper, we show that a second-order difference in Shafranov shift or geometric coordinates, or a difference between local and global profile implementations can bring a factor of two or more discrepancy in real frequency and growth rate. This suggests that an accurate global equilibrium is required for validation of gyrokinetic KBM simulations.

  4. Equilibrium and Instability near the Separatrix in Tokamaks

    NASA Astrophysics Data System (ADS)

    Zhou, Deng

    2016-10-01

    A local equilibrium model near the separatrix with up-down symmetric double nulls is developed in this work. Shaping features like elongation, triangularity and Shafranov shift of the reference magnetic surface are taken into account in the model. The poloidal magnetic field is determined as done in Ref.. The model can be used to study the effect of separatrix on the localized plasma modes such as peeling-ballooning modes near the tokamak edge region. As the first application we use it to calculate the Mercier index determining interchange modes.

  5. Nonlinear saturation of ballooning modes in tokamaks and stellarators

    PubMed Central

    Bauer, F.; Garabedian, P.; Betancourt, O.

    1988-01-01

    The spectral code BETAS computes plasma equilibrium in a toroidal magnetic field B = [unk]s × [unk]Ψ with remarkable accuracy because the finite difference scheme employed in the radial direction allows for discontinuities of the flux function Ψ across the nested surfaces s = const. Instability of higher modes in stellarators like the Heliotron E can be detected in roughly an hour on the best supercomputers by calculating bifurcated equilibria that are defined over just one field period. The method has been validated by comparing results about nonlinear saturation of ballooning modes in tokamaks with numerical data from the PEST code. PMID:16593984

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

    DOEpatents

    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.

  7. Tokamak with liquid metal for inducing toroidal electrical field

    DOEpatents

    Ohkawa, Tihiro

    1981-01-01

    A tokamak apparatus includes a vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within said vessel defines a toroidal space within the liner confines gas therein. Liquid metal fills the reservoir outside the liner. A magnetic field is established in the liquid metal to develop magnetic flux linking the toroidal space. The gas is ionized. The liquid metal and the toroidal space are moved relative to one another transversely of the space to generate electric current in the ionized gas in the toroidal space about its major axis and thereby heat plasma developed in the toroidal space.

  8. Tokamak with mechanical compression of toroidal magnetic field

    DOEpatents

    Ohkawa, Tihiro

    1981-01-01

    A tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A collapsible toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. A toroidal magnetic field is developed within the toroidal space about the major axis thereof. A toroidal plasma is developed within the toroidal space about the major axis thereof. Pressure is applied to the liquid metal to collapse the liner and reduce the volume of the toroidal space, thereby increasing the toroidal magnetic flux density therein.

  9. Impact of poloidal convective cells on momentum flux in tokamaks

    NASA Astrophysics Data System (ADS)

    Garbet, X.; Asahi, Y.; Donnel, P.; Ehrlacher, C.; Dif-Pradalier, G.; Ghendrih, P.; Grandgirard, V.; Sarazin, Y.

    2017-01-01

    Radial fluxes of parallel momentum due to E× B and magnetic drifts are shown to be correlated in tokamak plasmas. This correlation comes from the onset of poloidal convective cells generated by turbulence. The entire process requires a symmetry breaking mechanism, e.g. a mean shear flow. An analytical calculation shows that anti-correlation between the poloidal and parallel components of the turbulent Reynolds stress results in anti-correlation of the fluxes of parallel momentum generated by E× B and curvature drifts.

  10. Continuum kinetic modeling of the tokamak plasma edge

    SciTech Connect

    Dorf, M. A.; Dorr, M.; Rognlien, T.; Hittinger, J.; Cohen, R.

    2016-03-10

    In this study, the first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasmatransport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalous radial transport.

  11. EXPLICT CALULATIONS OF HOMOCLINIC TANGLES SURROUNDING MAGNETIC ISLANDS IN TOKAMAKS

    SciTech Connect

    ROEDER, R.K.W.; RAPOPORT, B.I.; EVANS, T.E.

    2002-06-01

    We present explicit calculations of the complicated geometric objects known as homoclinic tangles that surround magnetic islands in the Poincare mapping of a tokamak's magnetic field. These tangles are shown to exist generically in the magnetic field of all toroidal confinement systems. The geometry of these tangles provides an explanation for the stochasticity known to occur near the X-points of the Poincare mapping. Furthermore, the intersection of homoclinic tangles from different resonances provides an explicit mechanism for the non-diffusive transport of magnetic field lines between these resonance layers.

  12. Evidence of inward toroidal momentum convection in the JET tokamak.

    PubMed

    Tala, T; Zastrow, K-D; Ferreira, J; Mantica, P; Naulin, V; Peeters, A G; Tardini, G; Brix, M; Corrigan, G; Giroud, C; Strintzi, D

    2009-02-20

    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 and profile shape of the momentum diffusivity are similar to those of the ion heat diffusivity. A significant inward momentum pinch, up to 20 m/s, has been found. Both results are consistent with gyrokinetic simulations. This evidence is complemented in plasmas with internal transport barriers.

  13. Halo current diagnostic system of experimental advanced superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Chen, D. L.; Shen, B.; Granetz, R. S.; Sun, Y.; Qian, J. P.; Wang, Y.; Xiao, B. J.

    2015-10-01

    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.

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

    SciTech Connect

    Tala, T.; Zastrow, K.-D.; Brix, M.; Corrigan, G.; Giroud, C.; Naulin, V.; Peeters, A. G.; Tardini, G.; Strintzi, D.

    2009-02-20

    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 and profile shape of the momentum diffusivity are similar to those of the ion heat diffusivity. A significant inward momentum pinch, up to 20 m/s, has been found. Both results are consistent with gyrokinetic simulations. This evidence is complemented in plasmas with internal transport barriers.

  15. Continuum kinetic modeling of the tokamak plasma edge

    NASA Astrophysics Data System (ADS)

    Dorf, M. A.; Dorr, M. R.; Hittinger, J. A.; Cohen, R. H.; Rognlien, T. D.

    2016-05-01

    The first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalous radial transport.

  16. Tritium Experience in Large Tokamaks: Application to ITER

    SciTech Connect

    Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G.; Haanges, R.

    1998-05-01

    Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R{ampersand}D avenues to address them. This paper summarizes the presentations, discussions, and recommendations.

  17. Gyrokinetic simulation of isotope scaling in tokamak plasmas

    SciTech Connect

    Lee, W.W.; Santoro, R.A.

    1995-07-01

    A three-dimensional global gyrokinetic particle code in toroidal geometry has been used for investigating the transport properties of ion temperature gradient (ITG) drift instabilities in tokamak plasmas. Using the isotopes of hydrogen (H{sup +}), deuterium (D{sup +}) and tritium (T{sup +}), we have found that, under otherwise identical conditions, there exists a favorable isotope scaling for the ion thermal diffusivity, i.e., Xi decreases with mass. Such a scaling, which exists both at the saturation of the instability and also at the nonlinear steady state, can be understood from the resulting wavenumber and frequency spectra.

  18. Lower hybrid current drive in the PLT tokamak

    SciTech Connect

    Bernabei, S.; Daughney, C.; Efthimion, P.

    1982-07-01

    Order of magnitude improvements in the level and duration of current driven by lower hybrid waves have been achieved in the PLT tokamak. Steady currents up to 175 kA have been maintained for three seconds and 400 kA for 0.3 sec by the rf power alone. The principal current carrier appears to be a high energy (approx. 100 keV) electron component, concentrated in the central 20 to 40 cm diameter core of the 80 cm PLT discharge.

  19. Spectra of heliumlike krypton from Tokamak Fusion Test Reactor plasmas

    SciTech Connect

    Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M. ); Beiersdorfer, P.; Osterheld, A. ); Smith, A. ); Fraenkel, B. )

    1993-08-16

    Experiments were conducted on TFTR to study the radiation of krypton which will be important for future tokamaks, such as ITER, for the diagnostic of the central ion temperature and for the control of the energy release from the plasma by radiative cooling. The total krypton radiation was monitored, and satellite spectra of Kr XXXV were recorded with a high-resolution crystal spectrometer. Radiative cooling and reduced particle recycling at the plasma edge region were observed, in reasonable agreement with modeling calculations which included radial transport.

  20. Plasma rotation from momentum transport by neutrals in tokamaks

    NASA Astrophysics Data System (ADS)

    Omotani, J.; Pusztai, I.; Newton, S.; Fülöp, T.

    2016-12-01

    Neutral atoms can strongly influence the intrinsic rotation and radial electric field at the tokamak edge. Here, we present a framework to investigate these effects when the neutrals dominate the momentum transport. We explore the parameter space numerically, using highly flexible model geometries and a state of the art kinetic solver. We find that the most important parameters controlling the toroidal rotation and electric field are the major radius where the neutrals are localized and the plasma collisionality. This offers a means to influence the rotation and electric field by, for example, varying the radial position of the X-point to change the major radius of the neutral peak.

  1. Radiation−condensation instability in tokamaks with mixed impurities

    SciTech Connect

    Morozov, D. Kh.; Pshenov, A. A.

    2015-08-15

    Radiation−condensation instability (RCI) is one of the possible mechanisms behind the formation of microfaceted asymmetric radiation from the edge (MARFE) of a tokamak. It has been previously shown by the authors that RCI in carbon-seeded plasma can be stabilized using neon injection. Recently, beryllium- and tungsten-seeded plasmas became a subject of great interest. Therefore, in the present paper, RCI stability analysis of the edge plasma seeded with beryllium, tungsten, nitrogen, and carbon is performed. The influence of neutral hydrogen fluxes from the wall on the marginal stability limit is studied as well.

  2. Halo current diagnostic system of experimental advanced superconducting tokamak

    SciTech Connect

    Chen, D. L.; Shen, B.; Sun, Y.; Qian, J. P. Wang, Y.; Xiao, B. J.; Granetz, R. S.

    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.

  3. Deuterium and tritium separation in a tokamak reactor divertor layer

    NASA Astrophysics Data System (ADS)

    Tokar', M. Z.

    1989-04-01

    It's shown that the plasma isotope composition in a tokamak reactor divertor layer changes along the magnetic field and can notable differ from the gas composition in a pumping chamber. Heavier tritium must concentrate in the hot plasma far from the divertor plate due to thermal force stipulated by mutial collisions of deuterium and tritium ions. This circumstance is favourable from the point of view of tritium cycle optimization and must facilitate solution of the problem of tritium accumulation in the reactor construction elements.

  4. Neutral pumping rates for a next step tokamak ignition device

    SciTech Connect

    Galambos, J.D.; Peng, Y.K.M.; Heifetz, D.

    1985-01-01

    Neutral pumping rates are calculated for pump-limiter and divertor options of a next step tokamak ignition device using a method that accounts for the coupled effects of neutral transport and plasma transport. For both pump limiters and divertors the plasma flow into the channel surrounding the neutralizer plate is greatly reduced by the neutral recycling. The fraction of this flow that is pumped can be large (>50%) but in general is dependent on the particular geometry and plasma conditions. It is estimated that pumping speeds greater than or approximately 10/sup 5/ L/s are adequate for the exhaust requirements in the pump-limiter and the divertor cases.

  5. Diamagnetic thresholds for sawtooth cycling in tokamak plasmas

    SciTech Connect

    Halpern, Federico D.; Luetjens, Hinrich; Luciani, Jean-Francois

    2011-10-15

    The cycling dynamics of the internal kink mode, which drives sawtooth oscillations in tokamak plasmas, is studied using the three dimensional, non-linear magnetohydrodynamic (MHD) code XTOR-2F [H. Luetjens and J.-F. Luciani, J. Comput. Phys. 229, 8130 (2010)]. It is found that sawtooth cycling, which is characterized by quiescent ramps and fast crashes in the experiment, can be recovered in two-fluid MHD provided that a criterion of diamagnetic stabilization is fulfilled. The simulation results indicate that diamagnetic effects alone may be sufficient to drive sawteeth with complete magnetic reconnection in high temperature Ohmic plasmas.

  6. Electron cyclotron heating experiments on the DIII-D tokamak

    SciTech Connect

    Prater, R.; Austin, M.E.; Bernabei, S.

    1998-01-01

    Initial experiments on heating and current drive using second harmonic electron cyclotron heating (ECH) are being performed on the DIII-D tokamak using the new 110 GHz ECH system. Modulation of the ECH power in the frequency range 50 to 300 Hz and detection of the temperature perturbation by ECE diagnostics is used to validate the location of the heating. This technique also determines an upper bound on the width of the deposition profile. Analysis of electron cyclotron current drive indicates that up to 0.17 MA of central current is driven, resulting in a negative loop voltage near the axis.

  7. NASTRAN analysis of Tokamak vacuum vessel using interactive graphics

    NASA Technical Reports Server (NTRS)

    Miller, A.; Badrian, M.

    1978-01-01

    Isoparametric quadrilateral and triangular elements were used to represent the vacuum vessel shell structure. For toroidally symmetric loadings, MPCs were employed across model boundaries and rigid format 24 was invoked. Nonsymmetric loadings required the use of the cyclic symmetry analysis available with rigid format 49. NASTRAN served as an important analysis tool in the Tokamak design effort by providing a reliable means for assessing structural integrity. Interactive graphics were employed in the finite element model generation and in the post-processing of results. It was felt that model generation and checkout with interactive graphics reduced the modelling effort and debugging man-hours significantly.

  8. Turbulence studies in Tokamak boundary plasmas with realistic divertor geometry

    SciTech Connect

    Xu, X.Q.

    1998-10-14

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT [1] and the linearized shooting code BAL[2] to study turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant, resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters.

  9. Controlling tokamak geometry with three-dimensional magnetic perturbations

    SciTech Connect

    Bird, T. M.; Hegna, C. C.

    2014-10-15

    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.

  10. Dynamic modeling of transport and positional control of tokamaks

    SciTech Connect

    Jardin, S.C.; Pomphrey, N.; DeLucia, J.

    1985-10-01

    We describe here a numerical model of a free boundary axisymmetric tokamak plasma and its associated control systems. The plasma is modeled with a hybrid method using two-dimensional velocity and flux functions with surface-averaged MHD equations describing the evolution of the adiabatic invariants. Equations are solved for the external circuits and for the effects of eddy currents in nearby conductors. The method is verified by application to several test problems and used to simulate the formation of a bean-shaped plasma in the PBX experiment.

  11. JET, the largest tokamak on the eve of DT operation

    NASA Astrophysics Data System (ADS)

    Horton, L. D.

    2016-11-01

    The Joint European Torus (JET) is the world's largest operating tokamak and the only such machine capable of operating with the fuel mixture (deuterium-tritium) foreseen for a fusion reactor. Since it came into operation in 1983, JET has explored fusion plasmas "in conditions and dimensions approaching those of a fusion reactor" [1]. JET has demonstrated world-record levels of fusion power and energy production, in conditions where the ratio of the fusion power generated to the input power to the plasma, Q, approaches unity.

  12. Threshold condition for nonlinear tearing modes in tokamaks

    SciTech Connect

    Zabiego, M.F.; Callen, J.D.

    1996-03-01

    Low-mode-number tearing, mode nonlinear evolution is analyzed emphasizing the need for a threshold condition, to account for observations in tokamaks. The discussion is illustrated by two models recently introduced in the literature. The models can be compared with the available data and/or serve as a basis for planning some experiments in order to either test theory (by means of beta-limit scaling laws, as proposed in this paper) or attempt to control undesirable tearing modes. Introducing a threshold condition in the tearing mode stability analysis is found to reveal some bifurcation points and thus domains of intrinsic stability in the island dynamics operational space.

  13. Tritium experience in the Tokamak Fusion Test Reactor

    SciTech Connect

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Brooks, J.N.; Hogan, J.

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  14. Formation and Stability of Impurity "snakes" in Tokamak Plasmas

    SciTech Connect

    L. Delgado-Aparicio, et. al.

    2013-01-28

    New observations of the formation and dynamics of long-lived impurity-induced helical "snake" modes in tokamak plasmas have recently been carried-out on Alcator C-Mod. The snakes form as an asymmetry in the impurity ion density that undergoes a seamless transition from a small helically displaced density to a large crescent-shaped helical structure inside q < 1, with a regularly sawtoothing core. The observations show that the conditions for the formation and persistence of a snake cannot be explained by plasma pressure alone. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature

  15. Tokamak Transmutation of (nuclear) Waste (TTW): Parametric studies

    NASA Astrophysics Data System (ADS)

    Cheng, E. T.; Krakowski, R. A.; Peng, Y. K. M.

    Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses.

  16. Commissioning activities of the initial magnetic diagnostics for KSTAR tokamak

    NASA Astrophysics Data System (ADS)

    Lee, Sang Gon; Gyo Bak, Jun; Mie Ka, Eun

    2007-11-01

    The initial magnetic diagnostics for the KSTAR superconducting tokamak including three Rogowski coils, five flux/voltage loops, and sixty-four magnetic field probes have been successfully installed. The Rogowski coils, flux/voltage loops, and magnetic field probes measure the total plasma current, poloidal flux and loop voltage, and local poloidal magnetic field for the plasma position control and equilibrium studies, respectively. Accurate position measurements after installation for all of these initial magnetic diagnostics and in situ calibration for the Rogowski coils were finished. Data acquisition systems for these initial magnetic diagnostics are currently under preparation. Detail commissioning activities before the first plasma from these initial magnetic diagnostics will be presented.

  17. Enhancement of the Bootstrap Current in a Tokamak Pedestal

    NASA Astrophysics Data System (ADS)

    Kagan, Grigory; Catto, Peter J.

    2010-07-01

    The strong radial electric field in a subsonic tokamak pedestal modifies the neoclassical ion parallel flow velocity, as well as the radial ion heat flux. Existing experimental evidence of the resulting alteration in the poloidal flow of a trace impurity is discussed. We then demonstrate that the modified parallel ion flow can noticeably enhance the pedestal bootstrap current when the background ions are in the banana regime. Only the coefficient of the ion temperature gradient drive term is affected. The revised expression for the pedestal bootstrap current is presented. The prescription for inserting the modification into any existing banana regime bootstrap current expression is given.

  18. Oxygen impurity radiation from Tokamak-like plasmas

    NASA Technical Reports Server (NTRS)

    Rogerson, J. E.; Davis, J.; Jacobs, V. L.

    1977-01-01

    We have constructed a nonhydrodynamic coronal model for calculating radiation from impurity atoms in a heated plasma. Some recent developments in the calculation of dielectronic recombination rate coefficients and collisional excitation rate coefficients are included. The model is applied to oxygen impurity radiation during the first few milliseconds of a TFR Tokamak plasma discharge, and good agreement with experimental results is obtained. Estimates of total line and continuum radiation from the oxygen impurity are given. It is shown that impurity radiation represents a considerable energy loss.

  19. Halo current diagnostic system of experimental advanced superconducting tokamak.

    PubMed

    Chen, D L; Shen, B; Granetz, R S; Sun, Y; Qian, J P; Wang, Y; Xiao, B J

    2015-10-01

    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.

  20. Applying the new HIT results to tokamak and solar plasmas

    NASA Astrophysics Data System (ADS)

    Jarboe, Thomas; Sutherland, Derek; Hossack, Aaron; Nelson, Brian; Morgan, Kyle; Chris, Hansen; Benedett, Thomas; Everson, Chris; Penna, James

    2016-10-01

    Understanding sustainment of stable equilibria with helicity injection in HIT-SI has led to a simple picture of several tokamak features. Perturbations cause a viscous-like force on the current that flattens the λ profile, which sustains and stabilizes the equilibrium. An explanation of the mechanism is based on two properties of stable, ideal, two-fluid, magnetized plasma. First, the electron fluid is frozen to magnetic fields and, therefore, current flow is also magnetic field flow. Second, for a stable equilibrium the structure perpendicular to the flux surface resists deformation. Thus toroidal current is from electrons frozen in nested, rotating resilient flux surfaces. Only symmetric flux surfaces allow free differential current flow. Perturbations cause interference of the flux surfaces. Thus, perturbations cause forces that oppose differential electron rotation and forced differential flow produces a symmetrizing force against perturbations and instability. This mechanism can explain the level of field error that spoils tokamak performance and the rate of poloidal flux loss in argon-induced disruptions in DIII-D. This new understanding has led to an explanation of the source of the solar magnetic fields and the power source for the chromosphere, solar wind and corona. Please place in spheromak and FRC section with other HIT posters.

  1. Transient getter scheme for the Tokamak Fusion Test Reactor

    SciTech Connect

    Cecchi, J.L.; Cohen, S.A.; Sredniawski, J.J.

    1980-01-01

    The ability of the Tokamak Fusion Test Reactor (TFTR) to attain the largest fusion power gain depends critically on minimizing plasma contamination and controlling the densities of the reacting deuterium and tritium. Experiments on a number of tokamaks have demonstrated that gettering over an appreciable surface area (greater than or equal to 10%) of the vacuum vessel greatly facilitates both of these objectives. One particular problem in implementing a surface pumping system in TFTR, however, is a restriction on the maximum allowable tritium content of the getter. This restriction could require regeneration of the absorbed tritium after as few as 50 machine pulses. We have developed a scheme utilizing SAES Zr/Al getter modules which obviates the need for such frequent interruptions of machine operation by taking advantage of the pulsed operation of TFTR. With the Zr/Al getter at temperatures between 500/sup 0/C to 600/sup 0/C it is possible to achieve a quasi-steady state in the tritium loading where the quantity of tritium desorbed between pulses is equal to the quantity which is absorbed during a pulse. Since frequent thermal cycling is not required, this scheme also reduces the possibility of Zr/Al getter material fatigue.

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

    NASA Astrophysics Data System (ADS)

    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.

  3. Electron cyclotron emission measurements on the Texas Experimental Tokamak

    SciTech Connect

    Austin, M.E. Jr.

    1992-01-01

    A ten-channel grating polychromator was designed, constructed, and installed on the Texas Experimental Tokamak to monitor the second harmonic electron cyclotron emission. Electron temperature profiles were derived from measurements of the optically thick radiation for a variety of plasma confinement experiments. The radial and temporal evolution of T[sub e] has been characterized for electron cyclotron heated discharges with 150 kW of 60 GHz power. Comparisons were made of the heating efficiency of two type of ECH launchers. A focussed launcher was shown to have slightly better heating efficiency than an unfocussed launcher; however, the focussed antenna did not yield significantly higher electron temperatures as expected. A study of the time evolution of the electron temperature indicated that increased sawtooth activity limited the effectiveness of the focussed launcher. A focussing hog-horn antenna was fabricated and installed on the inboard side of the tokamak to measure emission directed towards the high-field side during ECH. Comparison of the radiation temperature profiles from low-field side and high-field side antennas indicates the creation of a nonthermal electron distribution by the heating. The results of the experiment compare favorably with theoretical predictions from a quasi-linear Fokker-Planck code of a 6 keV nonthermal population with a density about 1 percent of the thermal density.

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

    SciTech Connect

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

    2016-06-15

    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.

  5. Continuum Kinetic Modeling of the Tokamak Plasma Edge

    NASA Astrophysics Data System (ADS)

    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.

  6. Comments on experimental results of energy confinement of tokamak plasmas

    SciTech Connect

    Chu, T.K.

    1989-04-01

    The results of energy-confinement experiments on steady-state tokamak plasmas are examined. For plasmas with auxiliary heating, an analysis based on the heat diffusion equation is used to define heat confinement time (the incremental energy confinement time). For ohmically sustained plasmas, experiments show that the onset of the saturation regime of energy confinement, marfeing, detachment, and disruption are marked by distinct values of the parameter /bar n//sub e///bar j/. The confinement results of the two types of experiments can be described by a single surface in 3-dimensional space spanned by the plasma energy, the heating power, and the plasma density: the incremental energy confinement time /tau//sub inc/ = ..delta..W/..delta..P is the correct concept for describing results of heat confinement in a heating experiment; the commonly used energy confinement time defined by /tau//sub E/ = W/P is not. A further examination shows that the change of edge parameters, as characterized by the change of the effective collision frequency ..nu../sub e/*, governs the change of confinement properties. The totality of the results of tokamak experiments on energy confinement appears to support a hypothesis that energy transport is determined by the preservation of the pressure gradient scale length. 70 refs., 6 figs., 1 tab.

  7. Full f gyrokinetic method for particle simulation of tokamak transport

    SciTech Connect

    Heikkinen, J.A. Janhunen, S.J.; Kiviniemi, T.P.; Ogando, F.

    2008-05-10

    A gyrokinetic particle-in-cell approach with direct implicit construction of the coefficient matrix of the Poisson equation from ion polarization and electron parallel nonlinearity is described and applied in global electrostatic toroidal plasma transport simulations. The method is applicable for calculation of the evolution of particle distribution function f including as special cases strong plasma pressure profile evolution by transport and formation of neoclassical flows. This is made feasible by full f formulation and by recording the charge density changes due to the ion polarization drift and electron acceleration along the local magnetic field while particles are advanced. The code has been validated against the linear predictions of the unstable ion temperature gradient mode growth rates and frequencies. Convergence and saturation in both turbulent and neoclassical limit of the ion heat conductivity is obtained with numerical noise well suppressed by a sufficiently large number of simulation particles. A first global full f validation of the neoclassical radial electric field in the presence of turbulence for a heated collisional tokamak plasma is obtained. At high Mach number (M{sub p}{approx}1) of the poloidal flow, the radial electric field is significantly enhanced over the standard neoclassical prediction. The neoclassical radial electric field together with the related GAM oscillations is found to regulate the turbulent heat and particle diffusion levels particularly strongly in a large aspect ratio tokamak at low plasma current.

  8. Overview of the EUROfusion Medium Size Tokamak scientific program

    NASA Astrophysics Data System (ADS)

    Martin, Piero; Coda, Stefano; Eich, Thomas; Hakola, Antti; Meyer, Hendrik; EUROfusion MST1 Team; AUG Team; MAST-U Team; TCV Team

    2016-10-01

    The EUROfusion MST (Medium Size Tokamaks) task force is in charge of 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 2015/16 campaign in AUG and TCV and the future plans. We will discuss, among others, successful disruption and runaway electron control experiments with MGI and 3D fields, the achievement of full ELM suppression with RMP accompanied by the understanding of plasma response and the heat load pattern study, the exploration of regimes with impurity seeding at high P/R with 85% radiation fraction and good confinement, the study of tungsten fuzz, where W samples with pre-formed nanostructures were exposed to H-mode Helium plasmas and the investigation on advanced divertor concepts. A survey of MHD limits and of MHD control in standard and high-beta regimes will be presented. The results from the AUG campaign dedicated to He plasmas in support of ITER initial operation will also be presented, as well as analysis of old MAST data that reveal interesting features in the filamentary transport. See http://www.euro-fusionscipub.org/mst1.

  9. Transient getter scheme for the Tokamak Fusion Test Reactor

    SciTech Connect

    Cecchi, J.L.; Cohen, S.A.; Sredniawski, J.J.

    1980-01-01

    The ability of the Tokamak Fusion Test Reactor (TFTR) to attain the largest fusion power gain depends critically on minimizing plasma contamination and controlling the densities of the reacting deuterium and tritium. Experiments on a number of tokamaks have demonstrated that gettering over an appreciable surface area (> or = 10%) of the vacuum vessel greatly facilitates both of these objectives. One particular problem in implementing a surface pumping system in TFTR, however, is a restriction on the maximum allowable tritium content of the getter. This restriction could require regeneration of the absorbed tritium after as few as 50 machine pulses. We have developed a scheme utilizing SAES Zr/Al getter modules which obviates the need for such frequent interruptions of machine operation by taking advantage of the pulsed operation of TFTR. With the Zr/Al getter at temperatures between 500/sup 0/--600/sup 0/C it is possible to achieve a quasisteady state in the tritium loading where the quantity of tritium desorbed between pulses is equal to the quantity which is absorbed during a pulse. Since frequent thermal cycling is not required, this scheme also reduces the possibility of Zr/Al getter material fatigue.

  10. Numerical study of tokamak equilibria with arbitrary flow

    NASA Astrophysics Data System (ADS)

    Guazzotto, L.; Betti, R.; Manickam, J.; Kaye, S.

    2004-02-01

    The effects of toroidal and poloidal flows on the equilibrium of tokamak plasmas are numerically investigated using the code FLOW. The code is used to determine the changes in the profiles induced by large toroidal flows on NSTX-like equilibria [with NSTX being the National Spherical Torus Experiment, M. Ono, S.M. Kaye, Y.-K.M. Peng et al., Nucl. Fusion 40, 557 (2000)] where flows exceeding the sound speed lead to a considerable outward shift of the plasma. The code is also used to study the effects of poloidal flow when the flow velocity profile varies from subsonic to supersonic with respect to the poloidal sound speed. It is found that pressure and density profiles develop a pedestal structure characterized by radial discontinuities at the transonic surface where the poloidal velocity abruptly jumps from subsonic to supersonic values. These results confirm the conclusions of the analytic theory of R. Betti and J. P. Freidberg [Phys. Plasmas 7, 2439 (2000)], derived for a low-β, large aspect ratio tokamak with a circular cross section.

  11. Shock and soliton structures induced by poloidal flow in tokamaks

    NASA Astrophysics Data System (ADS)

    Taniuti, T.; Watanabe, K.; Ishii, Y.; Wakatani, M.

    1991-04-01

    When poloidal flow velocity, Up, becomes close to epsilon C sub s in tokamaks, fluid motion inside the flux surface is governed by a forced KdV equation in the small dissipation limit, where epsilon is an inverse aspect ratio and C sub s is a sound velocity. This implies that a stationary soliton structure appears at the inside region of toroidal plasma or theta is approx. pi, where theta is a poloidal angle. In the limit of large dissipation a forced Burgers equation is more appropriate. This case corresponds to an appearance of shock structure. When an averaged poloidal flow velocity is very close to epsilon C sub s, a shock may be seen in the outer region of toroidal plasma or theta is approx. 0. With the increase of the absolute vakue of (Up-epsilon C sub s) the shock position moves from theta is approx. 0 to theta is approx. pi. Recent tokamak experiments show the existence of poloidal flow with Up is approx. epsilon C sub s in the transition phase from L mode to H mode. Implications of the results in the experiment are discussed.

  12. Nearly axisymmetric hot plasmas in a highly rippled tokamak

    NASA Astrophysics Data System (ADS)

    Bellan, Paul

    2002-11-01

    Tokamak ohmic heating current flowing along toroidally rippled flux surfaces results in a poloidal torque. Since pressure gradients cannot offset torques, the torque drives plasma flows which convect plasma toroidally from ripple necks (high B_pol^2) to ripple bulges (low B_pol^2). Stagnation of the oppositely directed toroidal flows at the ripple bulges thermalizes the directed flow velocity ˜ B_pol/μ_0ρ , giving β _pol ˜1. These flows also convect frozen-in poloidal field lines which accumulate at the bulges enhancing the pinch force there and so reducing the bulge. Thus, a nearly axisymmetric β_pol ˜1 equilibrium is achieved using only a few TF coils. Particles bouncing in step between approaching flows will be Fermi accelerated to form a high energy tail. The ST tokamak magnetic mountain experiment [1] showed that, compared to a 1.8% ripple configuration, a 28% ripple configuration had four times the neutron production, and only a modest degradation of overall confinement; the former is consistent with the notion of Fermi acceleration of particles bouncing between colliding toroidal flows and the latter is consistent with ripple reduction due to toroidal convection of poloidal field lines. [1] W. Stodiek et al, Proc. 4th Intl. Conf. Plasma Phys. and Contr. Nuc. Fusion Res., (Madison, 1971), Vol. 1, p. 465

  13. Multichannel interferometer/polarimeter system for the RTP tokamak

    NASA Astrophysics Data System (ADS)

    van Lammeren, A. C. A. P.; Kim, S. K.; Donné, A. J. H.

    1990-10-01

    A nine-channel interferometer/polarimeter system is designed for the RTP tokamak (Rijnhuizen Tokamak Project, a =0.16 m, R =0.72 m, BT =2.5 T, Ip ≤ 200 kA, and plasma pulse duration 200 ms). A CO2-pumped dual-cavity FIR laser system is used to obtain two FIR laser beams of λ =432 μm with a frequency difference of 1 MHz. The FIR beams are expanded by a set of parabolic mirrors. Corner-cube mixers with Schottky diodes are used as detectors. The mixers are mounted on a linear rail system, such that their positions can be easily changed. The interferometer can be operated with a maximum of 19 channels, and will be extended with a polarimeter to measure the poloidal magnetic field distribution. It will be tried to increase the accuracy of the polarimeter by modulating the polarization of the incoming beam. A scheme for polarization modulation which is based on the idea of Dodel and Kunz will be presented in this paper. This scheme has the advantage that only one set of detectors is needed to measure the electron density and current density. The complete system will be presented along with some pilot experiments concerning the polarization modulation.

  14. Finite pressure effects on the tokamak sawtooth crash

    SciTech Connect

    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.

  15. Design of geometric phase measurement in EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Lan, Ting; Liu, Haiqing; Liu, Jian; Qin, Hong

    2016-10-01

    The aim of this work is to propose the optimum scheme for geometric phase measurement in EAST Tokamak. On the one hand, the experimental observation of geometric phase in plasma systems is an essential verification of the geometric phase theory by a new experimental technique. On the other hand, the measurement of geometric phase confirms geometric effect as a new system error in the existing diagnostics. The geometric phase in Faraday rotation angle for linearly polarized electromagnetic waves propagating in non-uniform magnetized plasmas is a good candidate for the first identification of geometric phase in plasma. In this work, the theoretical values of geometric phase for the probe beams of EAST Polarimeter-Interferometer (POINT) system are calculated by path integration in parameter space. Several schemes are proposed for the measurement of the geometric phase in POINT system by amplifying the geometric phase and enhancing the diagnostic resolution. To reach the conditions of the designed scheme for geometric phase measurement, the feasibility of replacing individual retro reflectors (RRs) with retro reflector array (RRA) in POINT system is verified experimentally. Corresponding results are beneficial for geometric phase measurement in EAST Tokamak.

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

    SciTech Connect

    Gates, D. A.; Brennan, D. P.; Delgado-Aparicio, L.; White, R. B.

    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.

  17. Overview of the Pegasus Extremely Low-Aspect Ratio Tokamak

    NASA Astrophysics Data System (ADS)

    Fonck, R.; Garstka, G.; Intrator, T.; Lewicki, B.; Thorson, T.; Toonen, R.; Tritz, K. L.; White, B.; Winz, G.

    1996-11-01

    Pegasus is a new experiment designed to explore the potential of Extremely Low Aspect Ratio Tokamaks (ELART) at very high toroidal β. Ohmic induction for plasma startup will be followed by ohmic sustainment initially and noninductive RF current drive in the future. Plasma parameters are projected to be Ip <= 0.3 MA, Bt < 0.2T, <β> ≈ 5-40 % or higher, A=1.1-2, R=0.2-0.4 m, and P_RF <= 2MW. Goals of the program include: demonstrate high-β spherical tokamak operation in the near term; examine the stability, n=0 stability properties at high elongation and low- A, confinement and scaling characteristics at A <= 1.25; and extend high power ST operation to the extrema of A <= 1.1. Hollow current profiles should be accessible in Pegasus using a fast current ramp during formation plus off-axis FWCD in the longer term. Recent changes to the design include: increased vacuum vessel height to allow for divertor operation with an internal X-point plus increased accessible elongations (i.,e., κ <= 3.7 at A = 1.25); additional coils for X-point control; and elimination of toroidal gaps in favor of a resistive vacuum vessel. Initial operation will emphasize ohmic access to high- β, followed by high power RF heating.

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

    SciTech Connect

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    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.

  19. Theoretical Predictions for Kr Radiation in a Tokamak

    NASA Astrophysics Data System (ADS)

    Lippmann, S. I.; Fournier, K. B.; Goldstein, W. H.

    1996-11-01

    There is current interest in the idea of using impurity radiation to control the energy exhaust from a future tokamak reactor. A suitable radiator would be one which radiates most strongly in the edge and does not cause a serious fuel dilution problem. Recently, the Hebrew University Lawrence Livermore Atomic Code (HULLAC) was adapted to this problem by making a prediction of the Kr radiation in a TFTR-like background plasma^1. Kr was considered an interesting candidate since the standard Post-Jensen cooling curves indicate a relatively large amount of radiation per ion in the edge relative to the center. HULLAC predicts even much more radiation (up to ~ 10^2× in the tokamak edge) from Kr in the edge than does the average-ion model. A similar result has is found for molybdenum^2. In this poster we will try to determine the nature of this descrepancy; whether it is do to a different ionization-recombination rates being used, an effect of the impurity transport assumptions used, or to some other theoretical problem. This work Supported by DOE subcontract B310463 to LLNL. ^1S. Lippmann, et al. Procedings of the 11th APS Topical Conference on Atomic Processes in Plasmas, 1996, p B20. ^2K. Fournier, these proceedings.

  20. Experimental modeling of eddy currents and defections for tokamak limiters

    SciTech Connect

    Hua, T.Q.; Knott, M.J.; Turner, L.R.; Wehrle, R.B.

    1986-01-01

    In a tokamak fusion reactor or other magnetically confined fusion device, a rapid decay of the magnetic field due to a disrupting plasma current induces eddy currents in the surrounding structures. These include the limiters, blanket, first wall, and vacuum vessel. The eddy currents, through interaction with the applied toroidal and poloidal magnetic fields, produce large mechanical torques and forces that deflect the structural components. An important coupling effect exists between the dynamic behavior of the structure and the transient eddy current. This coupling occurs when, as it deflects, the component intercepts additional magnetic flux. The coupling between deflection and eddy current could reduce the peak current, deflection, and other electromagnetic effects to a level far less severe than would be predicted if coupling is disregarded. In this study, experiments were performed to investigate deflection, current, and material stress in cantilever beams with the Fusion ELectromagnetic Induction eXperiment (FELIX) at the Argonne National Laboratory. Since structures near the plasma are typically cantilevered, the beams provide a good model for the limiter blades of a tokamak fusion reactor. 2 figs.

  1. Simulation of plasma flow in the DIII-D Tokamak

    SciTech Connect

    Porter, G. D., LLNL

    1998-06-19

    The importance of the parallel flow of primary and impurity ions in the Scrape-Off layer (SOL) of divertor tokamaks has been recognized recently. Impurity accumulation on the closed flux surfaces is determined in part by their parallel flow in the SOL. In turn, the parallel transport of the impurity ions is determined in part by drag from the primary ion flow. Measurement of flow in the DIII-D tokamak has begun recently. We describe initial results of modeling plasma ion flow using the 2-D code UEDGE in this paper. We assume the impurity (carbon) arises from chemical and physical sputtering from the walls surrounding the DIII-D plasma. We include six charge states of carbon in our simulations. We make detailed compaison with a multitude of SOL plasma diagnostics, including the flow measurement, to verify the UEDGE physics model. We begin the paper with a brief description of the plasma and neutral models in the UEDGE code in Section 2. We then present initial results of flow simulations and compare them with experimental measurement in Section 3. We conclude with a discussion of the dominant physics processes identified in the modeling in Section 4.

  2. ADVANCES IN DUST DETECTION AND REMOVAL FOR TOKAMAKS

    SciTech Connect

    Campos, A.; Skinner, C.H.

    2009-01-01

    Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. In the tokamak environment, large particles or fi bers can fall on the electrostatic detector potentially causing a permanent short. An electrostatic dust detector developed in the laboratory is being applied to the National Spherical Torus Experiment (NSTX). We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments at atmospheric pressure with varying nozzle designs, backing pressures, puff durations and exit fl ow orientations have given an optimal confi guration that effectively removes particles from a 25 cm² area. Similar removal effi ciencies were observed under a vacuum base pressure of 1 mTorr. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tri-polar grid of fi ne interdigitated traces has been designed that generates an electrostatic traveling wave for conveying dust particles to a “drain.” First trials with only two working electrodes have shown particle motion in optical microscope images.

  3. An overview of results from the TCV tokamak

    NASA Astrophysics Data System (ADS)

    Goodman, T. P.; Ahmed, S. M.; Alberti, S.; Andrèbe, Y.; Angioni, C.; Appert, K.; Arnoux, G.; Behn, R.; Blanchard, P.; Bosshard, P.; Camenen, Y.; Chavan, R.; Coda, S.; Condrea, I.; Degeling, A.; Duval, B. P.; Etienne, P.; Fasel, D.; Fasoli, A.; Favez, J.-Y.; Furno, I.; Henderson, M.; Hofmann, F.; Hogge, J.-P.; Horacek, J.; Isoz, P.; Joye, B.; Karpushov, A.; Klimanov, I.; Lavanchy, P.; Lister, J. B.; Llobet, X.; Magnin, J.-C.; Manini, A.; Marlétaz, B.; Marmillod, P.; Martin, Y.; Martynov, An.; Mayor, J.-M.; Mlynar, J.; Moret, J.-M.; Nelson-Melby, E.; Nikkola, P.; Paris, P. J.; Perez, A.; Peysson, Y.; Pitts, R. A.; Pochelon, A.; Porte, L.; Raju, D.; Reimerdes, H.; Sauter, O.; Scarabosio, A.; Scavino, E.; Seo, S. H.; Siravo, U.; Sushkov, A.; Tonetti, G.; Tran, M. Q.; Weisen, H.; Wischmeier, M.; Zabolotsky, A.; Zhuang, G.

    2003-12-01

    The Tokamak à Configuration Variable (TCV) tokamak (R = 0.88 m, a < 0.25 m, B < 1.54 T) programme is based on flexible plasma shaping and heating for studies of confinement, transport, control and power exhaust. Recent advances in fully sustained off-axis electron cyclotron current drive (ECCD) scenarios have allowed the creation of plasmas with high bootstrap fraction, steady-state reversed central shear and an electron internal transport barrier. High elongation plasmas, kgr = 2.5, are produced at low normalized current using far off-axis electron cyclotron heating and ECCD to broaden the current profile. Third harmonic heating is used to heat the plasma centre where the second harmonic is in cut-off. Both second and third harmonic heating are used to heat H-mode plasmas, at the edge and centre, respectively. The ELM frequency is decreased by the additional power. In separate experiments, the ELM frequency can be affected by locking to an external perturbation current in the internal coils of TCV. Spatially resolved current profiles are measured at the inner and outer divertor targets by Langmuir probe arrays during ELMs. The strong, reasonably balanced currents are thought to be thermoelectric in origin.

  4. Turbulent-driven intrinsic rotation in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Barnes, Michael; Parra, Felix; Lee, Jungpyo; Belli, Emily; Nave, Filomena; White, Anne

    2013-10-01

    Tokamak plasmas are routinely observed to rotate even in the absence of an externally applied torque. This ``intrinsic'' rotation exhibits several robust features, including rotation reversals with varying plasma density and current and rotation peaking at the transition from low confinement to high confinement regimes. Conservation of toroidal angular momentum dictates that the intrinsic rotation is determined by momentum redistribution within the plasma, which is dominated by turbulent transport. The turbulent momentum transport, and thus the intrinsic rotation profile, is driven by formally small effects that are usually neglected. We present a gyrokinetic theory that makes use of the smallness of the poloidal to total magnetic field ratio to self-consistently include the dominant effects driving intrinsic turbulent momentum transport in tokamaks. These effects (including slow radial profile variation, slow poloidal turbulence variation, and diamagnetic corrections to the equilibrium Maxwellian) have now been implemented in the local, delta-f gyrokinetic code GS2. We describe important features of the numerical implementation and show numerical results on intrinsic momentum transport that are qualitatively consistent with experimental rotation reversals.

  5. Electron ripple injection concept for tokamak transport control

    NASA Astrophysics Data System (ADS)

    Choe, W.; Ono, M.; Chang, C. S.

    1996-02-01

    A non-intrusive method for inducing a radial electric field (Er) based on electron ripple injection (ERI) is under development by the Princeton CDX-U group. Since Er is known to play an important role in the L-H and H-VH mode transition, it is therefore important to develop a non-intrusive tool to control the Er profile in tokamak plasmas. The present technique utilizes externally-applied local magnetic ripple fields to trap electrons at the edge, allowing them to penetrate towards the plasma center via ∇B and curvature drifts, causing the flux surfaces to charge up negatively. Electron cyclotron resonance heating (ECRH) is utilized to increase the trapped population and the electron drift velocity by raising the perpendicular energy of trapped electrons. The temperature anisotropy of resonant electrons in a tokamak plasma is calculated in order to investigate effects of ECRH on electrons. Simulations using a guiding-center orbit model have been performed to understand the behavior of suprathermal electrons in the presence of ripple fields. Examples for CDX-U and ITER are given.

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

    NASA Astrophysics Data System (ADS)

    Gates, D. A.; Brennan, D. P.; Delgado-Aparicio, L.; White, R. B.

    2015-06-01

    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.

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

    SciTech Connect

    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.

  8. On resistive magnetohydrodynamic studies of sawtooth oscillations in tokamaks

    SciTech Connect

    Aydemir, A. Y. Kim, J. Y.; Park, B. H.; Seol, J.

    2015-03-15

    A fundamental requirement for the validity and accuracy of any large-scale computation is sufficiently well-resolved length and time scales relevant to the problem under study. Ironically, despite the enormous computational resources available today, poorly resolved length scales in sophisticated nonlinear calculations are not uncommon. Using the internal kink mode that is responsible for tokamak sawtooth oscillations as an example, consequences of not resolving in sufficient detail the linear and nonlinear layer widths of the resistive n = 1 mode and its nonlinear spectrum are examined. Poor radial and spectral resolution are shown to cause nonphysical, large-scale stochasticity that can be erroneously associated with a fast temperature collapse and sawtooth crash. With the assistance of a nonlinear mode coupling model, a sufficiently well-resolved toroidal spectrum is shown to require at least an order of magnitude more toroidal modes than is commonly used at dissipation levels relevant to today's tokamaks. A subgrid-scale model is introduced that helps with the spectral resolution problem by reducing the required number of degrees of freedom from that of a well-resolved direct numerical simulation.

  9. SPECE: a code for Electron Cyclotron Emission in tokamaks

    SciTech Connect

    Farina, D.; Figini, L.; Platania, P.; Sozzi, C.

    2008-03-12

    The code SPECE has been developed for the analysis of electron cyclotron emission (ECE) in a general tokamak equilibrium. The code solves the radiation transport equation along the ray trajectories in a tokamak plasma, in which magnetic equilibrium and plasma profiles are given either analytically or numerically, for a Maxwellian plasma or a non thermal plasma characterized by a distribution function that is the sum of drifting Maxwellian distributions. Ray trajectories are computed making use of the cold dispersion relation, while the absorption and emission coefficients are obtained solving the relevant fully relativistic dispersion relation valid at high electron temperature. The actual antenna pattern is simulated by means of a multi-rays calculation, and the spatial resolution of the ECE measurements is computed by means of an algorithm that takes properly into account the emission along each ray of the beam. Wall effects are introduced in the code by means of a heuristic model. Results of ECE simulations in a standard ITER scenario are presented.

  10. ECE RADIOMETER UPGRADE ON THE DIII-D TOKAMAK

    SciTech Connect

    AUSTIN, ME; LOHR, J

    2002-08-01

    OAK A271 ECE RADIOMETER UPGRADE ON THE DIII-D TOKAMAK. The electron cyclotron emission (ECE) heterodyne radiometer diagnostic on DIII-D has been upgraded with the addition of eight channels for a total of 40. The new, higher frequency channels allow measurements of electron temperature into the magnetic axis in discharges at maximum field, 2.15 T. The complete set now extends over the full usable range of second harmonic emission frequencies at 2.0 T covering radii from the outer edge inward to the location of third harmonic overlap on the high field side. Full coverage permits the measurement of heat pulses and magnetohydrodynamic (MHD) fluctuations on both sides of the magnetic axis. In addition, the symmetric measurements are used to fix the location of the magnetic axis in tokamak magnetic equilibrium reconstructions. Also, the new higher frequency channels have been used to determine central T{sub e} with good time resolution in low field, high density discharges using third harmonic ECE in the optically gray and optically thick regimes.

  11. Electron cyclotron emission spectrometry on the Tokamak a Configuration Variable

    SciTech Connect

    Klimanov, I.; Porte, L.; Alberti, S.; Blanchard, P.; Fasoli, A.; Goodman, T.P.

    2005-09-15

    Electron cyclotron emission (ECE) measurements are an important component of the diagnostic suite on the Tokamak a Configuration Variable (TCV) [F. Hoffman et al., Plasma Phys. Controlled Fusion 36, B277 (1994)]. A recently installed, 24-channel dual-conversion heterodyne radiometer covering the radio frequency range 65-100 GHz and viewing from the low-field side (LFS) of the tokamak greatly enhances the system and, in combination with an existing radiometer viewing from the high-field side (HFS), allows simultaneous measurements of emission from the HFS and LFS. In addition, the new radiometer has multiple lines of sight that can receive the emission perpendicular to the toroidal magnetic field as well as with a finite k{sub parallel} (wave vector parallel to magnetic field). Such flexibility allows the LFS radiometer to make standard measurements of thermal emission and nonstandard measurements of nonthermal, anisotropic emission. The toroidal line of sight allows access to overdense plasma via mode converted emission. The enhanced ECE diagnostic is described and examples of measurements made in various configurations are presented.

  12. Feasibility study of a fission-suppressed tokamak fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  13. POLOIDAL MAGNETIC FIELD TOPOLOGY FOR TOKAMAKS WITH CURRENT HOLES

    SciTech Connect

    Puerta, Julio; Martin, Pablo; Castro, Enrique

    2009-07-26

    The appearance of hole currents in tokamaks seems to be very important in plasma confinement and on-set of instabilities, and this paper is devoted to study the topology changes of poloidal magnetic fields in tokamaks. In order to determine these fields different models for current profiles can be considered. It seems to us, that one of the best analytic descriptions is given by V. Yavorskij et al., which has been chosen for the calculations here performed. Suitable analytic equations for the family of magnetic field surfaces with triangularity and Shafranov shift are written down here. The topology of the magnetic field determines the amount of trapped particles in the generalized mirror type magnetic field configurations. Here it is found that the number of maximums and minimums of Bp depends mainly on triangularity, but the pattern is also depending of the existence or not of hole currents. Our calculations allow comparing the topology of configurations of similar parameters, but with and without whole currents. These differences are study for configurations with equal ellipticity but changing the triangularity parameters. Positive and negative triangularities are considered and compared between them.

  14. Runaway electrons in a Tokamak: A free-electron maser

    SciTech Connect

    Kurzan, B.; Steuer, K.; Suttrop, W.

    1997-01-01

    In Ohmic divertor plasma discharges of the ASDEX Upgrade tokamak a small population of runaway electrons is purposely generated at the beginning of the discharge. About 1.5 s after the runaways` generation fluctuating emission in the microwave region with a very narrow bandwidth is observed. The radiation can be explained by relativistic runaway electrons forming a free-electron maser in the tokamak: The dynamics of the runaways is simulated by taking into account the acceleration in the electric field, collisions with the plasma particles, synchrotron radiation losses and the resonances between the gyromotion and harmonics of the magnetic ripple field. According to this the runaways reach a final energy determined by the ripple resonance mechanism and are monoenergetic enough to form a cyclotron autoresonance maser. The guiding of the radiation along the bent path of the runaways is accomplished by nearly resonant runaways around the amplifying runaways. The emission frequency depends on the final energy of the runaways and the observed bandwidth via the maser`s gain on their particle density. Analyzing the emitted fluctuating emission is thus a potentially new diagnostic for runaways. {copyright} {ital 1997 American Institute of Physics.}

  15. Ellipticity and triangularity effects in tokamak Alfven spectrum

    NASA Astrophysics Data System (ADS)

    Puerta, Julio; Martin, Pablo; Castro, Enrique; Valdeblanquez, Eder

    2006-10-01

    Plasma configurations with ellipticity and triangularity are usual in tokamak experiments. These plasmas can be studied using a new system of coordinates of recent publications. Here this method has been applied to study Alfven spectrum in axisymmetric tokamaks with different values of ellipticity and triangularity [1-3]. Previous authors have developed numerical methods to obtain the Alfven spectrum using the Shafranov-Solove'v equilibrium flux function where the parameter ellipticity is also included [3]. Here more general configurations are treated and compared with the results of these authors, as well as those derived for the geometric optics or WKBJ approximation. The Alfven wave dispersion relation is obtained by the linearization of the MHD equations around a stationary equilibrium and the results are obtained by numerical calculations. [1] P. Martin, M. G. Haines and E. Castro, Phys. Plasma 12, 082506 (2005) [2] L. L. Lao, S. P. Hishman and R. M. Wieland, Phys. Fluids 24, 1431 (1981); H. Weitzner's Appendix. [3] G. O. Ludwig, Plasma Phys. Controlled Fusion 37, 633 (1995) [4] S. Novo, M. N'uñez and J. Rojo, Phys. Fluids B 3, 2967 (1991)

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

    PubMed

    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.

  17. Fast reciprocating probe system on the EAST superconducting tokamak.

    PubMed

    Zhang, W; Chang, J F; Wan, B N; Xu, G S; Xiao, C J; Li, B; Xu, C S; Yan, N; Wang, L; Liu, S C; Jiang, M; Liu, P

    2010-11-01

    A new fast reciprocating probe system (FRPS) has been built and installed on the outer midplane of the EAST tokamak to investigate the profiles of the boundary plasma parameters such as electron density and temperature. The system consists of a two-stage motion drive mechanism: slow motion and fast motion. The fast motion is powered by a servo motor, which drives the probe horizontally up to 50 cm to scan the edge region of the EAST tokamak. The maximum velocity achieved is 2 m/s. High velocity and flexible control of the fast motion are the remarkable features of this FRPS. A specially designed connector installed at the front end of the probe shaft makes it easy to install or replace the probe head on FRPS. During the latest experimental campaign in the spring of 2010, a probe head with seven tips, including two tips for a Mach probe, has been used. An example is given for simultaneous profile measurements of the plasma temperature, plasma density, and the plasma flow velocity.

  18. Dynamic simulations of the cryogenic system of a tokamak

    NASA Astrophysics Data System (ADS)

    Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

    2015-12-01

    Power generation in the next decades could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to 4.4K. The plasma works cyclically and the coil system is subjected to pulsed heat load which has to be handled by the refrigerator. By smoothing the variable loads, the refrigerator capacity can be set close to the average power; optimizing investment and operational costs. Within the “Broader Approach agreement” related to ITER project, CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) is in charge of providing the cryogenic system for the Japanese tokamak (JT-60SA), that is currently under construction in Naka. The system has been designed to handle the pulsed heat loads. To prepare the acceptance tests of the cryogenic system foreseen in 2016, both dynamic modelling and experimental tests on a scaled down mock-up are of high interest for assessing pulsed load smoothing control. After explaining HELIOS (HElium Loop for hIgh lOad Smoothing) operating modes, a dynamic model is presented, with results on the pulsed heat load scenarios. All the simulations have been performed with EcosimPro® and the associated cryogenic library CRYOLIB.

  19. Magnetic Diagnostics for the Lithium Tokamak eXperiment

    SciTech Connect

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

    2008-06-20

    The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with R0 = 0.4m, a = 0.26m, BTF ~ 3.4kG, IP ~ 400kA, and pulse length ~ 0.25s. 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.

  20. Nonlinear Fishbone Dynamics in Spherical Tokamaks with Toroidal Rotation

    NASA Astrophysics Data System (ADS)

    Wang, Feng; Fu, G. Y.

    2015-11-01

    Fishbone is ubiquitous in tokamak plasmas with fast ions. A numerical study of nonlinear dynamics of fishbone has been carried out in this work. Realistic parameters of NSTX are used to understand instability and nonlinear frequency chirping in tokamak plasmas. First, the effects of shear toroidal rotation are considered for fishbone instability. It's shown that with low qmin, it has small effects on the mode; while with high qmin, a new unstable region with a strong ballooning feature in mode structure appears. Second, a detailed study of nonlinear frequency chirping and energetic particles' dynamics is carried out. Linearly, the mode is driven by both trapped and passing particles, with dresonance condition ωd ~= ω for trapped particles and ωϕ +ωθ ~= ω for passing particles. As the mode grows, resonance particles oscillate and move outward in Pϕ space, which reduces particles' frequency. We believe that this is the main reason for the mode frequency chirping down. Finally, as the mode frequency chirping down, particles with lower orbit frequencies, which are non-resonant linearly, can turn into resonant particles in the nonlinear regime. This effect can sustain a quasi-steady state mode amplitude.