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Sample records for aditya tokamak

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

    Indian Academy of Sciences (India)

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

    2000-11-01

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as the discharge progresses. It is observed that during the current rise phase, current perturbation undergoes transition from = 5 poloidal structure to = 4 and then to = 3. At the time of current termination, = 2 perturbation is observed. It is observed that the mode frequency remains nearly constant (≈10 kHz) when poloidal mode structure changes from = 4 to = 2. This may be either an indication of mode coupling or a consequences of changes in the plasma electron temperature and density scale length.

  2. Overview of recent experimental results from the Aditya tokamak

    Science.gov (United States)

    Tanna, R. L.; Ghosh, J.; Chattopadhyay, P. K.; Raj, Harshita; Patel, Sharvil; Dhyani, P.; Gupta, C. N.; Jadeja, K. A.; Patel, K. M.; Bhatt, S. B.; Panchal, V. K.; Patel, N. C.; Chavda, Chhaya; Praveenlal, E. V.; Shah, K. S.; Makawana, M. N.; Jha, S. K.; Gopalkrishana, M. V.; Tahiliani, K.; Sangwan, Deepak; Raju, D.; Nagora, Umesh; Pathak, S. K.; Atrey, P. K.; Purohit, S.; Raval, J.; Joisa, Y. S.; Rao, C. V. S.; Chowdhuri, M. B.; Banerjee, S.; Ramaiya, N.; Manchanda, R.; Thomas, J.; Kumar, Ajai; Ajay, Kumar; Sharma, P. K.; Kulkarni, S. V.; Sathyanarayana, K.; Shukla, B. K.; Das, Amita; Jha, R.; Saxena, Y. C.; Sen, A.; Kaw, P. K.; Bora, D.; the ADITYA Team

    2017-10-01

    Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks, including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of a maximum plasma current of ~160 kA and discharge duration beyond ~250 ms with a plasma current flattop duration of ~140 ms have been obtained for the first time in ADITYA. The reproducibility of the discharge reproducibility has been improved considerably with lithium wall conditioning, and improved plasma discharges are obtained by precisely controlling the position of the plasma. In these discharges, chord-averaged electron density ~3.0-4.0  ×  1019 m-3 using multiple hydrogen gas puffs, with a temperature of the order of ~500-700 eV, have been achieved. Novel experiments related to disruption control are carried out and disruptions, induced by hydrogen gas puffing, are successfully mitigated using the biased electrode and ion cyclotron resonance pulse techniques. Runaway electrons are successfully mitigated by applying a short local vertical field (LVF) pulse. A thorough disruption database has been generated by identifying the different categories of disruption. Detailed analysis of several hundred disrupted discharges showed that the current quench time is inversely proportional to the q edge. Apart from this, for volt-sec recovery during the plasma formation phase, low loop voltage start-up and current ramp-up experiments have been carried out using electron cyclotron resonance heating (ECRH). Successful recovery of volt-sec leads to the achievement of longer plasma discharge durations. In addition, the neon gas puff assisted radiative improved confinement mode has also been achieved in ADITYA. All of the above mentioned experiments will be discussed in this paper.

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

    Indian Academy of Sciences (India)

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

    2015-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-15

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

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

    Science.gov (United States)

    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.

  6. 3D Monte-Carlo study of toroidally discontinuous limiter SOL configurations of Aditya tokamak

    Science.gov (United States)

    Sahoo, Bibhu Prasad; Sharma, Devendra; Jha, Ratneshwar; Feng, Yühe

    2017-08-01

    The plasma-neutral transport in the scrape-off layer (SOL) region formed by toroidally discontinuous limiters deviates from usual uniform SOL approximations when 3D effects caused by limiter discreteness begin to dominate. In an upgrade version of the Aditya tokamak, originally having a toroidally localized poloidal ring-like limiter, the newer outboard block and inboard belt limiters are expected to have smaller connection lengths and a multiple fold toroidal periodicity. The characteristics of plasma discharges may accordingly vary from the original observations of large diffusivity, and a net improvement and the stability of the discharges are desired. The estimations related to 3D effects in the ring limiter plasma transport are also expected to be modified and are updated by predictive simulations of transport in the new block limiter configuration. A comparison between the ring limiter results and those from new simulations with block limiter SOL shows that for the grids produced using same core plasma equilibrium, the modified SOL plasma flows and flux components have enhanced poloidal periodicity in the block limiter case. These SOL modifications result in a reduced net recycling for the equivalent edge density values. Predictions are also made about the relative level of the diffusive transport and its impact on the factors limiting the operational regime.

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

    Science.gov (United States)

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

    2014-11-01

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

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

    Science.gov (United States)

    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.

  9. Ion cyclotron resonance heating system on Aditya

    Indian Academy of Sciences (India)

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

    2005-02-01

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

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

    Science.gov (United States)

    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

  11. Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meglicki, Z

    1995-09-19

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

  12. Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meglicki, Z

    1995-09-19

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

  13. Operation and control of high power Gyrotrons for ECRH systems in SST-1 and Aditya

    Energy Technology Data Exchange (ETDEWEB)

    Shukla, B.K., E-mail: shukla@ipr.res.in; Bora, D.; Jha, R.; Patel, Jatin; Patel, Harshida; Babu, Rajan; Dhorajiya, Pragnesh; Dalakoti, Shefali; Purohit, Dharmesh

    2016-11-15

    Highlights: • Operation and control of high power Gyrotrons. • Data acquisition and control (DAQ) for Gyrotron system. • Ignitron based crowbar protection. • VME and PXI based systems. - Abstract: The Electron Cyclotron Resonance Heating (ECRH) system is an important heating system for the reliable start-up of tokamak. The 42 GHz and 82.6 GHz ECRH systems are used in tokamaks SST-1 and Aditya to carry out ECRH related experiments. The Gyrotrons are high power microwave tubes used as a source for ECRH systems. The Gyrotron is a delicate microwave tube, which deliver megawatt level power at very high voltage ∼40–50 kV with the current requirement ∼10 A–50 A. The Gyrotrons are associated with the subsystems like: High voltage power supplies (Beam voltage and anode voltage), dedicated crowbar system, magnet, filament and ion pump power supplies, cooling, interlocks and a dedicated data acquisition & control (DAC) system. There are two levels of interlocks used for the protection of Gyrotron: fast interlocks (arcing, beam over current, dI/dt, anode voltage and anode over current etc.) operate within 10 μs and slow interlocks (cooling, filament, silence of Gyrotron, ion pump and magnet currents) operate within 100 ms. Two Gyrotrons (42 GHz/500 kW/500 ms and 82.6 GHz/200 kW/1000 s) have been commissioned on dummy load for full parameters. The 42 GHz ECRH system has been integrated with SST-1 & Aditya tokamak and various experiments have been carried out related to ECRH assisted breakdown and start-up of tokamak at fundamental and second harmonic. These Gyrotrons are operated with VME based data acquisition and control (DAC) system. The DAC system is capable to acquire 64 digital and 32 analog signals. The system is used to monitor & acquire the data and also used for slow interlocks for the protection of Gyrotron. The data acquired from the system are stored online on VME system and after the shot stored in a file in binary format. The MDSPlus, a set of

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

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Ramesh; Singh, Manoj; Jadav, H M; Misra, Kishor; Kulkarni, S V, E-mail: rjoshi@ipr.res.i [Institute for plasma research, Bhat, Gandhinagar- 382428 (India)

    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 twice, 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

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

    Science.gov (United States)

    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

  16. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

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

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

    Indian Academy of Sciences (India)

    M Jagadeeshwari; J Govindarajan

    2000-11-01

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

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

    Indian Academy of Sciences (India)

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

    2000-11-01

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

  19. Status of tokamak research

    Energy Technology Data Exchange (ETDEWEB)

    Rawls, J.M. (ed.)

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design. (MOW)

  20. Tokamak Systems Code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

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

  2. Texas Experimental Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1998-01-01

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

  4. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1996-01-01

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

  5. Transport in gyrokinetic tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Mynick, H.E.; Parker, S.E.

    1995-01-01

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

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

    NARCIS (Netherlands)

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

    1997-01-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines.

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

    NARCIS (Netherlands)

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

    1997-01-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. S

  8. Magnetic confinement experiment -- 1: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1994-12-31

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

  9. Edge turbulence in tokamaks

    Science.gov (United States)

    Nedospasov, A. V.

    1992-12-01

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

  10. Dust Measurements in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-23

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

  11. Alcator C-Mod Tokamak

    Data.gov (United States)

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

  12. The ETE spherical Tokamak project

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  13. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-12-01

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

  14. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  15. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  16. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

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

  17. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

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

  18. A compact Tokamak transmutation reactor

    Institute of Scientific and Technical Information of China (English)

    QiuLi-Jian; XiaoBing-Jia

    1997-01-01

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

  19. Bootstrap Current in Spherical Tokamaks

    Institute of Scientific and Technical Information of China (English)

    王中天; 王龙

    2003-01-01

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

  20. Cryogenic needs for future tokamaks

    Science.gov (United States)

    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.

  1. Options for an ignited tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.

    1984-02-01

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

  2. Magnetic confinement experiment. I: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1995-08-01

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

  3. The role of limiter in Egyptor Tokamak

    CERN Document Server

    Ei-Sisi, A B

    2002-01-01

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

  4. Linear optimal control of tokamak fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Anomalous particle pinch in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Miskane, F.; Garbet, X. [Association Euratom-CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France); Dezairi, A.; Saifaoui, D. [Faculte des Sciences Ain Chok, Casablanca (Morocco)

    2000-06-01

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

  6. OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Simulation of dust statistical characteristics in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, R.D.; Pigarov, A.Yu.; Krasheninnikov, S.I.; Rosenberg, M.; Mendis, D.A. [University of California, San Diego, La Jolla, California, 92093 (United States)

    2008-03-15

    In this work we analyze the size (radius) distribution function of dust particles in tokamak plasmas during a steady state discharge. A relation between the radius distribution function of dust in the plasma and the radius distribution of dust injected from tokamak walls is obtained using a Green's function formalism. Numerical simulations of the dust radius distribution function in a tokamak plasma with the Dust Transport (DUSTT) code are used to obtain the analytical form of the Green's function semi-empirically. It is demonstrated that the Green's function obtained can be used to predict qualitatively the dust size distributions in the tokamak plasmas. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  8. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  9. Plasma diagnostics using synchrotron radiation in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  10. D-D tokamak reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Economic considerations of commercial tokamak options

    Energy Technology Data Exchange (ETDEWEB)

    Dabiri, A.E.

    1986-05-01

    Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m/sup 2/, which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m/sup 2/, will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e).

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

    CERN Document Server

    Ghim, Young-chul

    2013-01-01

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

  13. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

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

  14. Microtearing modes in tokamak discharges

    Energy Technology Data Exchange (ETDEWEB)

    Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Weiland, J. [Department of Applied Physics, Chalmers University, S41296 Gothenburg (Sweden); Luo, L. [IBM Research, Oak Ridge, Tennessee 37831 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80308 (United States)

    2016-06-15

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

  15. Microtearing modes in tokamak discharges

    Science.gov (United States)

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

    2016-06-01

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

  16. Up-down asymmetric tokamaks

    CERN Document Server

    Ball, Justin

    2016-01-01

    Bulk toroidal rotation has proven capable of stabilising both dangerous MHD modes and turbulence. In this thesis, we explore a method to drive rotation in large tokamaks: up-down asymmetry in the magnetic equilibrium. We seek to maximise this rotation by finding optimal up-down asymmetric flux surface shapes. First, we use the ideal MHD model to show that low order external shaping (e.g. elongation) is best for creating up-down asymmetric flux surfaces throughout the device. Then, we calculate realistic up-down asymmetric equilibria for input into nonlinear gyrokinetic turbulence analysis. Analytic gyrokinetics shows that, in the limit of fast shaping effects, a poloidal tilt of the flux surface shaping has little effect on turbulent transport. Since up-down symmetric surfaces do not transport momentum, this invariance to tilt implies that devices with mirror symmetry about any line in the poloidal plane will drive minimal rotation. Accordingly, further analytic investigation suggests that non-mirror symmetri...

  17. Global gyrokinetic simulation of tokamak transport

    Energy Technology Data Exchange (ETDEWEB)

    Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T. [Univ. of Texas, Austin, TX (United States). Inst. for Fusion Studies]|[Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)

    1998-10-01

    A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or {eta}{sub i}({eta}{sub i} {equivalent_to} {partial_derivative}{ell}nT{sub i}/{partial_derivative}{ell}n n{sub i}) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling.

  18. Physics of Tokamak Plasma Start-up

    Science.gov (United States)

    Mueller, Dennis

    2012-10-01

    This tutorial describes and reviews the state-of-art in tokamak plasma start-up and its importance to next step devices such as ITER, a Fusion Nuclear Science Facility and a Tokamak/ST demo. Tokamak plasma start-up includes breakdown of the initial gas, ramp-up of the plasma current to its final value and the control of plasma parameters during those phases. Tokamaks rely on an inductive component, typically a central solenoid, which has enabled attainment of high performance levels that has enabled the construction of the ITER device. Optimizing the inductive start-up phase continues to be an area of active research, especially in regards to achieving ITER scenarios. A new generation of superconducting tokamaks, EAST and KSTAR, experiments on DIII-D and operation with JET's ITER-like wall are contributing towards this effort. Inductive start-up relies on transformer action to generate a toroidal loop voltage and successful start-up is determined by gas breakdown, avalanche physics and plasma-wall interaction. The goal of achieving steady-sate tokamak operation has motivated interest in other methods for start-up that do not rely on the central solenoid. These include Coaxial Helicity Injection, outer poloidal field coil start-up, and point source helicity injection, which have achieved 200, 150 and 100 kA respectively of toroidal current on closed flux surfaces. Other methods including merging reconnection startup and Electron Bernstein Wave (EBW) plasma start-up are being studied on various devices. EBW start-up generates a directed electron channel due to wave particle interaction physics while the other methods mentioned rely on magnetic helicity injection and magnetic reconnection which are being modeled and understood using NIMROD code simulations.

  19. The Spherical Tokamak MEDUSA for Mexico

    Science.gov (United States)

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

    2011-10-01

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

  20. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

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

  1. Overview of spherical tokamak research in Japan

    Science.gov (United States)

    Takase, Y.; Ejiri, A.; Fujita, T.; Fukumoto, N.; Fukuyama, A.; Hanada, K.; Idei, H.; Nagata, M.; Ono, Y.; Tanaka, H.; Uchida, M.; Horiuchi, R.; Kamada, Y.; Kasahara, H.; Masuzaki, S.; Nagayama, Y.; Oishi, T.; Saito, K.; Takeiri, Y.; Tsuji-Iio, S.

    2017-10-01

    Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

  2. Tokamak Spectroscopy for X-Ray Astronomy

    Science.gov (United States)

    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.

  3. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

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

    2010-10-01

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

  4. Electron cyclotron emission diagnostics on KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  5. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.

    1985-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-13

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

  8. Tokamak startup with electron cyclotron heating

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-04-01

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

  9. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

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

  10. Tokamak Transport Studies Using Perturbation Analysis

    NARCIS (Netherlands)

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

    1990-01-01

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

  11. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

  12. Design method of divertor in tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Noriaki (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Itoh, Sanae; Tanaka, Masaaki; Itoh, Kimitaka

    1991-03-01

    Computational method to design the efficient divertor configuration in tokamak reactor is presented. The two-dimensional code has been developed to analyze the distributions of the plasma and neutral particles for realistic configurations. Using this code, a method to design the efficient divertor configuration is developed. An example of new divertor, which consists of the baffle and fin plates, is analyzed. (author).

  13. UCLA Tokamak Program Close Out Report.

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Robert John [UCLA/retired

    2014-02-04

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

  14. Tokamak Transport Studies Using Perturbation Analysis

    NARCIS (Netherlands)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-02-15

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

  16. Banana orbits in elliptic tokamaks with hole currents

    Science.gov (United States)

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

    2015-03-01

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

  17. First Divertor Operation on the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    YANG Qing-Wei; CAO Zeng; LI Xiao-Dong; MAO Wei-Cheng; ZHOU Cai-Pin; WANG En-Yao; YAN Jian-Cheng; LIU Yong; HL-2A team; DING Xuan-Tong; YAN Long-Wen; XUAN Wei-Min; LIU De-Quan; CHEN Liao-Yuan; SONG Xian-Ming; YUAN Bao-Shan; ZHANG Jin-Hua

    2004-01-01

    @@ HL-2A device is the first divertor tokamak in China. One of its main subjects is to study the features of the divertor plasma. In the last campaign, the first divertor configuration has been achieved and sustained on the HL-2A tokamak. Here we give a brief description about the HL-2A tokamak, diagnostics arrangements, and the equilibrium analysis results on divertor configuration. The main results of divertor experiments are also presented.

  18. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  19. Boundary Plasma Turbulence Simulations for Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  20. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

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

  1. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-01-01

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

  2. Self-Organized Stationary States of Tokamaks.

    Science.gov (United States)

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

    2015-11-20

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

  3. Neoclassical transport in high [beta] tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cowley, S.C.

    1992-12-01

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

  4. Resistive interchange instability in reversed shear tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  5. Axisymmetric instability in a noncircular tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lipschultz, B.

    1979-10-01

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

  6. Internal Kink Instability in Shaped Tokamaks

    Institute of Scientific and Technical Information of China (English)

    王中天; 王龙

    2002-01-01

    A criterion of an ideal internal kink mode is derived for a shaped tokamak configuration in which q-profile is very flat in the core region. A combining criterion is obtained including the necessary criterion of Mercier and the sufficient criterion of Lortz. The new criterion makes progress compared with the necessary criterion of Mercier. In the elongated plasma, a poloidal beta can cause instability, while the triangularity has a stabilizing effect. The result is applicable for DIII-D and SUNIST.

  7. EU Integrated Tokamak Modelling (ITM) Task Force

    Institute of Scientific and Technical Information of China (English)

    A Becoulet

    2007-01-01

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

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

    Science.gov (United States)

    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.

  9. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  10. Relativistic runaway electrons in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Jaspers, R.E.

    1995-02-03

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

  11. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

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

    2015-11-01

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

  12. Nonlinear Simulation Studies of Tokamaks and STs

    Energy Technology Data Exchange (ETDEWEB)

    W. Park; J. Breslau; J. Chen; G.Y. Fu; S.C. Jardin; S. Klasky; J. Menard; A. Pletzer; B.C. Stratton; D. Stutman; H.R. Strauss; L.E. Sugiyama

    2003-07-07

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

  13. SOL Width Scaling in the MAST Tokamak

    Science.gov (United States)

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

    2002-11-01

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

  14. Nondiffusive plasma transport at tokamak edge

    Science.gov (United States)

    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.

  15. A simulation study of a controlled tokamak plasma

    Science.gov (United States)

    Fujii, N.; Niwa, Y.

    1980-03-01

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

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

    NARCIS (Netherlands)

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

    1994-01-01

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

  17. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    Haverkort, J.W.

    2013-01-01

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

  18. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

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

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

    Indian Academy of Sciences (India)

    R Kumar; S K Saha

    2000-11-01

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

  20. System assessment of helical reactors in comparison with tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  2. Energetic particles in spherical tokamak plasmas

    Science.gov (United States)

    McClements, K. G.; Fredrickson, E. D.

    2017-05-01

    Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together with higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion-ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of beam ion

  3. Application of MDSplus on EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

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

    2007-01-01

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

  4. Differential and Integral Models of TOKAMAK

    Directory of Open Access Journals (Sweden)

    Ivo Dolezel

    2004-01-01

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

  5. Neutron skyshine calculations for the PDX tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, F.J.; Nigg, D.W.

    1979-01-01

    The Poloidal Divertor Experiment (PDX) at Princeton will be the first operating tokamak to require a substantial radiation shield. The PDX shielding includes a water-filled roof shield over the machine to reduce air scattering skyshine dose in the PDX control room and at the site boundary. During the design of this roof shield a unique method was developed to compute the neutron source emerging from the top of the roof shield for use in Monte Carlo skyshine calculations. The method is based on simple, one-dimensional calculations rather than multidimensional calculations, resulting in considerable savings in computer time and input preparation effort. This method is described.

  6. 3D MHD Simulations of Tokamak Disruptions

    Science.gov (United States)

    Woodruff, Simon; Stuber, James

    2014-10-01

    Two disruption scenarios are modeled numerically by use of the CORSICA 2D equilibrium and NIMROD 3D MHD codes. The work follows the simulations of pressure-driven modes in DIII-D and VDEs in ITER. The aim of the work is to provide starting points for simulation of tokamak disruption mitigation techniques currently in the CDR phase for ITER. Pressure-driven instability growth rates previously observed in simulations of DIIID are verified; Halo and Hiro currents produced during vertical displacements are observed in simulations of ITER with implementation of resistive walls in NIMROD. We discuss plans to exercise new code capabilities and validation.

  7. On circulating power of steady state tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, Kimitaka [National Inst. for Fusion Science, Nagoya (Japan); Itoh, Sanae; Fukuyama, Atsushi; Yagi, Masatoshi

    1996-03-01

    Circulating power for the sustenance and profile control of the steady state tokamak plasmas is discussed. The simultaneous fulfillment of the MHD stability at high beta value, the improved confinement and the stationary equilibrium requires the rotation drive as well as the current drive. In addition to the current drive efficiency, the efficiency for the rotation drive is investigated. The direct rotation drive by the external torque, such as the case of beam injection, is not efficient enough. The mechanism and the magnitude of the spontaneous plasma rotation are studied. (author)

  8. Dust divertor for a tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tang, X Z [Los Alamos National Laboratory; Delzanno, G L [Los Alamos National Laboratory

    2009-01-01

    Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.

  9. Development of atomic beam probe for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M., E-mail: bertam@sze.hu [Széchenyi István University, EURATOM Association, Győr (Hungary); Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, EURATOM Association, Budapest (Hungary); Havlícek, J.; Háček, P. [Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics (Czech Republic)

    2013-11-15

    Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies.

  10. Simulating W Impurity Transport in Tokamaks

    Science.gov (United States)

    Younkin, Timothy R.; Green, David L.; Lasa, Ane; Canik, John M.; Wirth, Brian D.

    2016-10-01

    The extreme heat and charged particle flux to plasma facing materials in magnetically confined fusion devices has motivated Tungsten experiments such as the ``W-Ring'' experiment on the DIII-D tokamak to investigate W divertor viability. In this domain, the transport of W impurities from their tile locations to other first-wall tiles is highly relevant to material lifetimes and tokamak operation. Here we present initial results from a simulation of this W transport. Given that sputtered impurities may experience prompt redeposition near the divertor strikepoint, or migrate far from its origin to the midplane, there is a need to track the global, 3-D, impurity redistribution. This is done by directly integrating the 6-D Lorentz equation of motion (plus thermal gradient terms and relevant Monte-Carlo operators) for the impurity ions and neutrals under background plasma parameters determined by the SOLPS edge plasma code. The geometric details of the plasma facing components are represented to a fidelity sufficient to examine the global impurity migration trends with initial work also presented on advanced surface meshing capabilities targeting high fidelity simulation. This work is supported by U.S. DOE Office of Science SciDAC project on plasma-surface interactions under US DOE contract DE-AC05-00OR22725.

  11. Electromagnetic simulations of tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  12. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.

    1996-07-01

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

  14. Systems assessment of the five Starlite tokamak power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G. [Los Alamos National Lab., NM (United States)

    1996-12-31

    The ARIES team has assessed the power-plant attractiveness of the following five tokamak physics regimes: (1) steady state, first stability regime; (2) pulsed, first stability regime; (3) steady state, second stability regime; (4) steady state, reversed shear; and (5) steady state, low aspect ratio. Cost-based systems analysis of these five tokamak physics regimes suggests that an electric power plant based upon a reversed-shear tokamak is significantly more economical than one based on any of the other four physics regimes. Details of this comparative systems analysis are described herein. 11 refs., 3 figs., 1 tab.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.

    1996-07-01

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

  16. Nonlinear stabilization of tokamak microturbulence by fast ions

    CERN Document Server

    Citrin, J; Garcia, J; Haverkort, J W; Hogeweij, G M D; Jenko, F; Johnson, T; Mantica, P; Pueschel, M J; Told, D; contributors, JET-EFDA

    2013-01-01

    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.

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

    Institute of Scientific and Technical Information of China (English)

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

    2005-01-01

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

  18. Power Deposition on Tokamak Plasma-Facing Components

    CERN Document Server

    Arter, Wayne; Fishpool, Geoff

    2014-01-01

    The SMARDDA software library is used to model plasma interaction with complex engineered surfaces. A simple flux-tube model of power deposition necessitates the following of magnetic fieldlines until they meet geometry taken from a CAD (Computer Aided Design) database. Application is made to 1) models of ITER tokamak limiter geometry and 2) MASTU tokamak divertor designs, illustrating the accuracy and effectiveness of SMARDDA, even in the presence of significant nonaxisymmetric ripple field. SMARDDA's ability to exchange data with CAD databases and its speed of execution also give it the potential for use directly in the design of tokamak plasma facing components.

  19. Development of a free boundary Tokamak Equilibrium Solver (TES) for Advanced Study of Tokamak Equilibria

    CERN Document Server

    Jeon, Y M

    2015-01-01

    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 after all in equilibrium calculation with a free-boundary 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 on variations of computational domain, the mathematical correctness and accuracy of equilibrium profiles are checked by a direct comparison with an analytic equilibrium known as a generalized Solovev equilibrium, and the governing force balance relation is tested by examining the intrinsic axisymmetric instabilities. As a valuable application, a snowflake equilibrium that requires a second order zero of the poloidal magnetic field is discussed in the circumstance of KSTAR coil system.

  20. Embedded data acquisition system with MDSPlus

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-15

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

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

    Indian Academy of Sciences (India)

    Krishnendu Bhattacharyya; N R Ray

    2000-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sa, Wanderley Pires de

    1996-12-31

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

  3. A review of ELMs in divertor tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Hill, D.N.

    1996-05-23

    This paper reviews what is known about edge localized modes (ELMs), with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M,JT-60U, and TCV tokamaks and are reported here. At least three types of ELMs have been identified and their salient features determined. Type-1 giant ELMs can cause the sudden loss of up to 10-15% of the plasma stored energy but their amplitude ({Delta}W/W) does not increase with increasing power. Type- 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1-3% of the stored energy). All ELMs increase the scrape- off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger that the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-Mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of Type-1 ELM amplitude and frequency have been measured in several tokamaks and has recently been applied to predictions of the ELM Size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with a very small type-3 ELMs can be maintained with heating power sell above the H-mode threshold, where giant type-1 ELMs can be maintained with heating power well above the H-mode threshold, where Giant type-1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H- mode discharges should not be overlooked.

  4. Toroidicity Dependence of Tokamak Edge Safety Factor and Shear

    Institute of Scientific and Technical Information of China (English)

    SHIBingren

    2002-01-01

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

  5. Compact Ignition Tokamak Program: status of FEDC studies

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, C.A.

    1985-01-01

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

  6. Tokamak Start-up under Assistance of RF Waves

    Institute of Scientific and Technical Information of China (English)

    方瑜德

    2004-01-01

    To improve the start-up behavior of tokamak discharges and realize the low loop voltage start-up is required by performance of large scale, full superconductor tokamaks. In recent years, some kinds of RF wave have been used to assist the start-up and some exciting results have been gained. This paper introduce the investigation on both in physical principle and experimental research of the start-up process, in which high frequency RF waves were used to assist it.

  7. NEOCLASSICAL TRANSPORT IN A TOKAMAK WITH ELECTRIC SHEAR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

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

  8. DIII-D Advanced Tokamak Research Overview

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

  9. Vertically stabilized elongated cross-section tokamak

    Science.gov (United States)

    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.

  10. Sliding Mode Control of a Tokamak Transformer

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-08

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

  11. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  12. Advantages of iron core in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

    1977-01-01

    A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti ..beta.. (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated.

  13. Dissipative nonlinear structures in tokamak plasmas

    Directory of Open Access Journals (Sweden)

    K. A. Razumova

    2001-01-01

    Full Text Available A lot of different kinds of instabilities may be developed in high temperature plasma located in a strong toroidal magnetic field (tokamak plasma. Nonlinear effects in the instability development result in plasma self-organization. Such plasma has a geometrically complicated configuration, consisting of the magnetic surfaces imbedded into each other and split into islands with various characteristic numbers of helical twisting. The self-consistency of the processes means that the transport coefficients in plasma do not depend just on the local parameters, being a function of the whole plasma configuration and of the forces affecting it. By disrupting the bonds between separate magnetic surfaces filled with islands, one can produce zones of reduced transport in the plasma, i.e. “internal thermal barriers”, allowing one essentially to increase the plasma temperature and density.

  14. Nonlinear lower hybrid modeling in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Napoli, F.; Schettini, G. [Università Roma Tre, Dipartimento di Ingegneria, Roma (Italy); Castaldo, C.; Cesario, R. [Associazione EURATOM/ENEA sulla Fusione, Centro Ricerche Frascati (Italy)

    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.

  15. A lithium deposition system for tokamak devices*

    Science.gov (United States)

    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

  16. Transport Bifurcation in a Rotating Tokamak Plasma

    CERN Document Server

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

    2010-01-01

    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 non-zero 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.

  17. General Tokamak Circuit Simulation Program-GTCSP

    Energy Technology Data Exchange (ETDEWEB)

    Matsukawa, Makoto; Miura, Yushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Aoyagi, Tetsuo

    1997-05-01

    General Tokamak Circuit Simulation Program (GTCSP) was originally developed for the design work of JT-60 Power Supply System in JAERI. Therefore the prepared models (components) to be analyzed are generator, thyristor converter and coils. This is one of the unique points of GTCSP in comparison with other conventional electric circuit analysis program, because they make a circuit from the small devices such as resister, coil, condenser, transistor and so on. However, GTCSP is also clearly conventional because it is possible to construct an electric circuit freely with the prepared components. Moreover, a similar function could be realized by addition a new component to GTCSP. This report is assumed to be used as an User Manual of the GTCSP, not only to present the development and the analytical functions. Then some useful examples are described, and how to get graphic outputs are also mentioned. (author)

  18. Safety factor profile control in a tokamak

    CERN Document Server

    Bribiesca Argomedo, Federico; Prieur, Christophe

    2014-01-01

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

  19. Resistive edge mode instability in stellarator and tokamak geometries

    Science.gov (United States)

    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.

  20. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  1. Trajectory planning of tokamak flexible in-vessel inspection robot

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hesheng [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Lai, Yinping; He, Tao [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China)

    2015-10-15

    Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  3. Basic Physics of Tokamak Transport Final Technical Report.

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Amiya K.

    2014-05-12

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

  4. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  5. Fast bolometric measurements on the TCV tokamak

    Science.gov (United States)

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

    1999-12-01

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

  6. Neoclassical transport in high {beta} tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cowley, S.C.

    1992-12-01

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

  7. Aspects of Tokamak toroidal magnet protection

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-07-01

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

  8. Zeeman Spectroscopy of Tokamak Edge Plasmas

    Science.gov (United States)

    Hey, J. D.; Chu, C. C.; Mertens, Ph.

    2002-12-01

    Zeeman spectroscopy is a valuable tool both for diagnostic purposes, and for more fundamental studies of atomic and molecular processes in the boundary region of magnetically confined fusion plasmas (B ≃ 1 to 10 T). The method works well when the Zeeman (Paschen-Back) effect plays an important, or dominant, rôle in relation to other broadening mechanisms (Doppler, Stark, resonant excitation transfer) in determining the spectral line shape. For impurity species identification and temperature determination, Zeeman spectroscopy has advantages over charge-exchange recombination spectroscopy from highly excited radiator states, since spectral features practically unique to the species under investigation are analysed. It also provides useful information on probable mechanisms of line production (e.g. sputtering mechanisms, electron impact-induced dissociative excitation from molecules in the edge plasma), and on the temperature evolution of lower charge states in the process of convection inwards or diffusion outwards from the hotter plasma interior. Where different physical processes are responsible for different sections of the line profile — especially in the case of hydrogen isotopes — Zeeman spectroscopy can provide a set of characteristic temperatures for each section. The method is introduced in both passive and active spectroscopy, and general principles of the Zeeman effect are discussed with special reference to régimes of interest for the tokamak. Relevant physical processes (sputtering mechanisms, electron impact-induced dissociative excitation from molecules in the edge plasma, and ion-atom collisional heating mechanisms) are illustrated by sample spectra.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-01-01

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

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

    Science.gov (United States)

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

    2011-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Valencia A, R

    2006-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

    Directory of Open Access Journals (Sweden)

    Jing Li

    2014-01-01

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

  15. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  16. Tokamak dust particle size and surface area measurement

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W.J.; Smolik, G.R.; Anderl, R.A.; Pawelko, R.J.; Hembree, P.B.

    1998-07-01

    The INEEL has analyzed a variety of dust samples from experimental tokamaks: General Atomics` DII-D, Massachusetts Institute of Technology`s Alcator CMOD, and Princeton`s TFTR. These dust samples were collected and analyzed because of the importance of dust to safety. The dust may contain tritium, be activated, be chemically toxic, and chemically reactive. The INEEL has carried out numerous characterization procedures on the samples yielding information useful both to tokamak designers and to safety researchers. Two different methods were used for particle characterization: optical microscopy (count based) and laser based volumetric diffraction (mass based). Surface area of the dust samples was measured using Brunauer, Emmett, and Teller, BET, a gas adsorption technique. The purpose of this paper is to present the correlation between the particle size measurements and the surface area measurements for tokamak dust.

  17. Hybrid Method for Tokamak MHD Equilibrium Configuration Reconstruction

    Institute of Scientific and Technical Information of China (English)

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

    2007-01-01

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

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

    Science.gov (United States)

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

    2016-10-01

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

  19. A CONCEPT FOR NEXT STEP ADVANCED TOKAMAK FUSION DEVICE

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  1. Resistive demountable toroidal-field coils for tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Measurement of Current Profile in a Tokamak Through AC Modulation

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    The plasma current is modulated with an alternating current (ac) component in a frequency range of 90 Hz~900 Hz in the plateau discharge phase in the CT-6B tokamak. A plasma electric conductivity profile in a form of (1 - r2/a2)α with a parameter α, which is fitted with the experimental data, can be determined. The effects of magnetic shear in a tokamak field configuration on the current penetration are taken into account in the numerical simulation. The measurement method and obtained results are discussed.

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

    Institute of Scientific and Technical Information of China (English)

    XIE Han; LIAO Ziying

    2008-01-01

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

  4. Computer simulation of transport driven current in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Nunan, W.J.; Dawson, J.M. (University of California at Los Angeles, Department of Physics, 405 Hilgard Avenue, Los Angeles, California 90024-1547 (United States))

    1994-09-19

    We have investigated transport driven current in tokamaks via 2+1/2 dimensional, electromagnetic, particle-in-cell simulations. These have demonstrated a steady increase of toroidal current in centrally fueled plasmas. Neoclassical theory predicts that the bootstrap current vanishes at large aspect ratio, but we see equal or greater current growth in straight cylindrical plasmas. These results indicate that a centrally fueled and heated tokamak may sustain its toroidal current, even without the seed current'' which the neoclassical bootstrap theory requires.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  6. Magnetic flux reconstruction methods for shaped tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Fusion neutron diagnostics on ITER tokamak

    Science.gov (United States)

    Bertalot, L.; Barnsley, R.; Direz, M. F.; Drevon, J. M.; Encheva, A.; Jakhar, S.; Kashchuk, Y.; Patel, K. M.; Arumugam, A. P.; Udintsev, V.; Walker, C.; Walsh, M.

    2012-04-01

    ITER is an experimental nuclear reactor, aiming to demonstrate the feasibility of nuclear fusion realization in order to use it as a new source of energy. ITER is a plasma device (tokamak type) which will be equipped with a set of plasma diagnostic tools to satisfy three key requirements: machine protection, plasma control and physics studies by measuring about 100 different parameters. ITER diagnostic equipment is integrated in several ports at upper, equatorial and divertor levels as well internally in many vacuum vessel locations. The Diagnostic Systems will be procured from ITER Members (Japan, Russia, India, United States, Japan, Korea and European Union) mainly with the supporting structures in the ports. The various diagnostics will be challenged by high nuclear radiation and electromagnetic fields as well by severe environmental conditions (ultra high vacuum, high thermal loads). Several neutron systems with different sensitivities are foreseen to measure ITER expected neutron emission from 1014 up to almost 1021 n/s. The measurement of total neutron emissivity is performed by means of Neutron Flux Monitors (NFM) installed in diagnostic ports and by Divertor Neutron Flux Monitors (DNFM) plus MicroFission Chambers (MFC) located inside the vacuum vessel. The neutron emission profile is measured with radial and vertical neutron cameras. Spectroscopy is accomplished with spectrometers looking particularly at 2.5 and 14 MeV neutron energy. Neutron Activation System (NAS), with irradiation ends inside the vacuum vessel, provide neutron yield data. A calibration strategy of the neutron diagnostics has been developed foreseeing in situ and cross calibration campaigns. An overview of ITER neutron diagnostic systems and of the associated challenging engineering and integration issues will be reported.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fraboulet, D.

    1996-09-17

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

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

    Institute of Scientific and Technical Information of China (English)

    SHIBingren; LONGYongxing; DONGJiaqi; LIWenzhong; JIAOYiming; WANGAike

    2002-01-01

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

  11. MHD analysis of edge instabilities in the JET tokamak

    NARCIS (Netherlands)

    Perez von Thun, Christian Pedro

    2004-01-01

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

  12. Disruption avoidance through active magnetic feedback in tokamak plasmas

    Science.gov (United States)

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

    2014-10-01

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

  13. Kazakhstan tokamak for material testing conceptual design and basic parameters

    Energy Technology Data Exchange (ETDEWEB)

    Korotkov, V.A. E-mail: korotkov@sintez.niiefa.spb.su; Azizov, E.A.; Cherepnin, Yu.S.; Dokouka, V.N.; Ya.Dvorkin, N.; Khayrutdinov, R.R.; Krylov, V.A.; Kuzmin, E.G.; Leykin, I.N.; Mineev, A.B.; Shkolnik, V.S.; Shestakov, V.P.; Shapovalov, G.V.; Tazhibaeva, I.L.; Tikhomirov, L.N.; Yagnov, V.A

    2001-10-01

    The construction of a special machine for plasma facing material testing under powerful and particle and heat flux deposition is necessary for progress of researches in the field of controlled fusion to industrial application. Kazakhstan tokamak for material testing (KTM) is planned as spherical tokamak with moderate-to-low aspect ratio (A=2) and high plasma and vacuum vessel elongation, that allows to reach high plasma parameters, large power-intensity at a compact arrangement of design elements and low requirements to a toroidal magnetic field. KTM tokamak is planned in order to investigate the following issues: (1) Plasma confinement in tokamak with A=2, plasma parameters and configurations working window; (2) Differed kinds of divertor plates under power flux of plasma to divertor volume; (3) Plasma-wall interaction (different materials and coating) and plasma-limiter configurations. In the paper the basic parameters of the machine are given. The design of magnet system with poloidal field coils, vacuum vessel and divertor are submitted.

  14. Dynamic diagnostics of the error fields in tokamaks

    Science.gov (United States)

    Pustovitov, V. D.

    2007-07-01

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

  15. Bulk Ion Heating with ICRF Waves in Tokamaks

    DEFF Research Database (Denmark)

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

    2015-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

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

    Science.gov (United States)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

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

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

    Institute of Scientific and Technical Information of China (English)

    ZHENGYongzhen

    2002-01-01

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

  19. General Description of Ideal Tokamak MHD Instability Ⅱ

    Institute of Scientific and Technical Information of China (English)

    石秉仁

    2002-01-01

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

  20. Test particle transport in perturbed magnetic fields in tokamaks

    NARCIS (Netherlands)

    de Rover, M.; Schilham, A.M.R.; Montvai, A.; Cardozo, N. J. L.

    1999-01-01

    Numerical calculations of magnetic field line trajectories in a tokamak are used to investigate the common hypotheses that (i) field lines in a chaotic field make a Gaussian random walk and (ii) that the poloidal component of the magnetic field is uniform in regions with a chaotic magnetic field. Bo

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

    DEFF Research Database (Denmark)

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

    2009-01-01

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

  2. Conceptual design of Remote Control System for EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  3. Tokamak Scenario Trajectory Optimization Using Fast Integrated Simulations

    Science.gov (United States)

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

    2015-11-01

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

  4. Current ramps in tokamaks: from present experiments to ITER scenarios

    NARCIS (Netherlands)

    Imbeaux, F.; Citrin, J.; Hobirk, J.; Hogeweij, G. M. D.; Kochl, F.; Leonov, V. M.; Miyamoto, S.; Nakamura, Y.; Parail, V.; Pereverzev, G.; Polevoi, A.; Voitsekhovitch, I.; Basiuk, V.; Budny, R.; Casper, T.; Fereira, J.; Fukuyama, A.; Garcia, J.; Gribov, Y. V.; Hayashi, N.; Honda, M.; Hutchinson, I. H.; Jackson, G.; Kavin, A. A.; Kessel, C. E.; Khayrutdinov, R. R.; Labate, C.; Litaudon, X.; Lomas, P. J.; Lonnroth, J.; Luce, T.; Lukash, V. E.; Mattei, M.; Mikkelsen, D.; Nunes, I.; Peysson, Y.; Politzer, P.; Schneider, M.; Sips, G.; Tardini, G.; Wolfe, S. M.; Zhogolev, V. E.

    2011-01-01

    In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for

  5. Pellet Enhanced Performance on the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    DING Xuan-Tong; LIU Yi; ZHOU Yan; PAN Yu-Dong; CUI Zheng-Ying; HUANG Yuan; LIU Ze-Tian; SHI Zhong-Bing; JI Xiao-Quan; XIAO Wei-Wen; LIU Yong; YANG Qing-Wei; YAN Long-Wen; ZHU Gen-Liang; XIAO Zheng-Gui; LIU De-Quan; CAO Zeng; GAO Qing-Di; LONG Yong-Xing

    2006-01-01

    @@ Enhanced confinement has been achieved by the centre fuelling of pellet injection on the HL-2A tokamak. The energy confinement time increases from 50ms to 140ms after the pellet injection. Experimental results show that the improvement of the confinement is related to the decrease of the electron heat transport.

  6. Sensitivity of transient synchrotron radiation to tokamak plasma parameters

    Energy Technology Data Exchange (ETDEWEB)

    Fisch, N.J.; Kritz, A.H.

    1988-12-01

    Synchrotron radiation from a hot plasma can inform on certain plasma parameters. The dependence on plasma parameters is particularly sensitive for the transient radiation response to a brief, deliberate, perturbation of hot plasma electrons. We investigate how such a radiation response can be used to diagnose a variety of plasma parameters in a tokamak. 18 refs., 13 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Edmonds, P.H. [Texas Univ., Austin, TX (United States). Fusion Research Center; Medley, S.S.; Young, K.M. [Princeton Univ., NJ (United States). Plasma Physics Lab.] [and others

    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.

  8. Ion Temperature-Measurements in Tokamak Plasmas by Rutherford Scattering

    NARCIS (Netherlands)

    Vanblokland, A. A. E.; Barbian, E. P.; Donne, A. J. H.; van der Grift, A. F.; Grimbergen, T. W. M.; Oyevaar, T.; Schüller, F. C.; Tammen, H. F.; Vanderven, H. W.; Vijverberg, T. F.; Dewinter, F. D. A.; Bertschinger, G.; Cosler, A.; Korten, M.

    1992-01-01

    A Rutherford scattering diagnostic has been applied at the TEXTOR tokamak to obtain spatially and temporally resolved information on the temperature of the bulk ions in the plasma. In the experimental setup, a helium atomic beam (30-keV, 12-mA equivalent current) passes vertically through the plasma

  9. Design and construction of Alborz tokamak vacuum vessel system

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-15

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

  10. LIDAR Thomson scattering for advanced tokamaks. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G. [and others

    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.

  11. Fueling studies on the lithium tokamak experiment

    Science.gov (United States)

    Lundberg, Daniel Patrick

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

  12. Vacuum system of SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India); Pathan, Firozkhan; George, Siju; Semwal, Pratibha; Dhanani, Kalpesh; Paravastu, Yuvakiran; Thankey, Prashant; Ramesh, Gattu; Himabindu, Manthena; Pradhan, Subrata [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India)

    2013-10-15

    Highlights: ► Air leaks developed during ongoing SST-1 cooldown campaign were detected online using RGA. ► The presence of N{sub 2} and O{sub 2} gases with the ratio of their partial pressures with ∼3.81:1 confirmed the air leaks. ► Baking of SST-1 was done efficiently by flowing hot N{sub 2} gas in C-channels welded on inner surfaces without any problem. ► In-house fabricated demountable bull nose couplers were demonstrated for high temperature and pressure applications. ► Cryopumping effect was observed when liquid helium cooled superconducting magnets reached below 63 K. -- Abstract: Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN{sub 2} cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10{sup −4} mbar and 1.0 × 10{sup −5} mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10{sup −6} mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10{sup −5} mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ∼3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10{sup −5} mbar is achieved inside the cryostat. Baking of the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, James D. [Auburn University, Auburn, Alabama; Anderson, D.T. [University of Wisconsin, Madison; Cianciosa, M. [Auburn University, Auburn, Alabama; Franz, P. [EURATOM / ENEA, Italy; Harris, J. H. [Oak Ridge National Laboratory (ORNL); Hartwell, G. H. [Auburn University, Auburn, Alabama; Hirshman, Steven Paul [ORNL; Knowlton, Stephen F. [Auburn University, Auburn, Alabama; Lao, Lang L. [General Atomics, San Diego; Lazarus, Edward Alan [ORNL; Marrelli, L. [Association EURATOM ENEA Fusion, Consorzio RFX, Padua, Italy; Maurer, D. A. [Auburn University, Auburn, Alabama; Schmitt, J. C. [Princeton Plasma Physics Laboratory (PPPL); Sontag, A. C. [Oak Ridge National Laboratory (ORNL); Stevenson, B. A. [Auburn University, Auburn, Alabama; Terranova, D. [Association EURATOM ENEA Fusion, Consorzio RFX, Padua, Italy

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bourdelle, C

    2000-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lohr, J.; Burrell, K.H. [General Atomics, San Diego, CA (United States); Coda, S. [Massachusetts Inst. of Tech., Cambridge, MA (United States)] [and others

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-05-01

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

  19. Multipoint Thomson scattering diagnostic for the ETE tokamak

    Science.gov (United States)

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

    Energy Technology Data Exchange (ETDEWEB)

    Zimin, A. M., E-mail: zimin@power.bmstu.ru [Bauman Moscow State Technical University (Russian Federation); Krupin, V. A. [National Research Centre Kurchatov Institute (Russian Federation); Troynov, V. I. [Bauman Moscow State Technical University (Russian Federation); Klyuchnikov, L. A. [National Research Centre Kurchatov Institute (Russian Federation)

    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.

  1. Overview of ARIES-RS tokamak fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Najmabadi, F. [California Univ., San Diego, CA (United States). Fusion Energy Research Program

    1998-09-01

    In order for fusion power to be widely accepted in the next century, it should offer advantages compared to available sources of energy. The Starlite study has examined the ability of tokamak-based power plants to compete with fusion energy sources. A set of top-level system requirements and goals for system economics, safety and waste disposal, and reliability and availability were established during extensive consultations with US electric utilities and industry representatives. Five different tokamak plasma operation modes were considered and different technology options (e.g. choice of structural material, coolant, breeder) were developed and assessed. Based on this assessment, the ARIES-RS design study was initiated to examine a power plant based on the reversed-shear mode of plasma operation, coupled to a fusion power core which uses high-performance lithium-cooled vanadium components. An overview of the ARIES-RS design is presented in this paper. (orig.) 14 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, Aditya K.; Ganesh, R., E-mail: ganesh@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar, 382428 (India); Brunner, S.; Vaclavik, J.; Villard, L. [CRPP, EPFL, 1015 Lausanne (Switzerland)

    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.

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

    Institute of Scientific and Technical Information of China (English)

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

    2007-01-01

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

  4. Stability and heating of a poloidal divertor tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Biddle, A. P.; Dexter, R. N.; Holly, D. T.; Lipschultz, B.; Osborne, T. H.; Prager, S. C.; Shepard, D.A., Sprott, J.C.; Witherspoon, F. D.

    1980-06-01

    Five experimental studies - two stability and three heating investigations - have been carried out on Tokapole II, a Tokamak with a four node poloidal divertor. First, discharges have been attained with safety factor q as low as 0.6 over most of the column without degradation of confinement, and correlation of helical instability onset with current profile shape is being studied. Second, the axisymmetric instability has been investigated in detail for various noncircular cross-sectional shapes, and results have been compared with a numerical stability code adapted to the Tokapole machine. Third, application of high power fast wave ion cyclotron resonance heating doubles the ion temperature and permits observation of heating as a function of harmonic number and spatial location of the resonance. Fourth, low power shear Alfven wave propagation is underway to test the applicability of this heating method to tokamaks. Fifth, preionization by electron cyclotron heating has been employed to reduce the startup loop voltage by approx. 60%.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-11-15

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

  6. Specification of asymmetric VDE loads of the ITER tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, C., E-mail: christian.bachmann@iter.org [ITER Organization CS90 046, 13067St. Paul lez Durance, Cedex (France); Sugihara, M.; Roccella, R.; Sannazzaro, G.; Gribov, Y. [ITER Organization CS90 046, 13067St. Paul lez Durance, Cedex (France); Riccardo, V.; Hender, T.C.; Gerasimov, S.N. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Pautasso, G. [Max Planck Institute for Plasma Physics, D-85748 Garching (Germany); Belov, A.; Lamzin, E. [D.V. Efremov Institute, Scientific Technical Centre ' Sintez' (Russian Federation); Roccella, M. [L. T. Calcoli, 23087 Merate, Lecco (Italy)

    2011-10-15

    During asymmetric vertical displacement events (AVDEs) associated with the kink mode of the plasma two asymmetry phenomena were observed in existing tokamaks, in particular in JET . The related halo currents flowing in the passive structure were identified as the cause of asymmetric EM loads on tokamak components. The first phenomenon is a toroidal peak of the poloidal halo current that flows in the passive structure. The second phenomenon is that the toroidal plasma current is not uniform toroidally, so a toroidally non-uniform current flows in the vessel . The specification of the expected characteristics of both phenomena as well as of the consequent asymmetric loads in ITER are summarized here. The related loads are specified for likely, unlikely and extremely unlikely AVDEs.

  7. Localized measurements of turbulence in the TORE SUPRA tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Devynck, P.; Garbet, X.; Laviron, C.; Payan, J.; Saha, S.K. (CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France)); Gervais, F.; Hennequin, P.; Quemeneur, A.; Truc, A. (Ecole Polytechnique, 91 - Palaiseau (France))

    1993-01-01

    A collective infra-red laser scattering diagnostic has been installed on the TORE SUPRA tokamak for the measurement of plasma density fluctuations. For the range of wavenumbers explored (3-15) cm[sup -1], the scattering angles are very weak ([approx] 1 mrad). Consequently, the scattering signals are averaged along the whole observation chord, resulting in poor longitudinal spatial localization. However, by virtue of the pitch angle variation of the magnetic field lines in the tokamak, and of the perpendicularity of the turbulence wavevector to these field lines, it has been possible to obtain partial spatial resolution along the direction of the beam. Good agreement between the experimental and theoretical angular resolution of the diagnostic as well as the results of cross-correlation performed on the signals obtained by two simultaneous probing beams also justify this novel concept. (Author).

  8. Analytical solutions for Tokamak equilibria with reversed toroidal current

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-15

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

  9. 3D MHD disruptions simulations of tokamaks plasmas

    Science.gov (United States)

    Paccagnella, Roberto; Strauss, Hank; Breslau, Joshua

    2008-11-01

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

  10. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-15

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2004-01-01

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

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

    Science.gov (United States)

    Moslehi-Fard, Mahmoud

    2013-02-01

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

  13. First Neutron Spectrometry Measurement at the HL-2A Tokamak

    CERN Document Server

    Xi, Yuan; Xufei, Xie; Zhongjing, Chen; Xingyu, Peng; Tieshuan, Fan; Jinxiang, Chen; Xiangqing, Li; Guoliang, Yuan; Jinwei, Yang; Qingwei, Yang

    2013-01-01

    A compact neutron spectrometer based on the liquid scintillator is presented for the neutron energy spectrum measurement at the HL-2A tokamak. The spectrometer has been well characterized and a fast digital pulse shape discrimination software has been developed using the charge comparison method. A digitizer data acquisition system with the maximum frequency of 1 MHz can work under the high count rate environment at HL-2A. Specific radiation shielding and magnetic shielding for the spectrometerhas been designed for the neutron spectrum measurement at the HL-2A Tokamak. For the analysis of the pulse height spectrum, dedicated numerical simulation utilizing NUBEAM combining with GENESIS has been made to obtain the neutron energy spectrum, following which the transportation process from the plasma to the detector has been evaluated with Monte Carlo calculations. The distorted neutron energy spectrum has been folded with response matrix of the liquid scintillation spectrometer, and good consistency has been found...

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

    Science.gov (United States)

    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/{{\

  15. Transition to subcritical turbulence in a tokamak plasma

    CERN Document Server

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

    2016-01-01

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

  16. Effects of the equilibrium model on impurity transport in tokamaks

    CERN Document Server

    Skyman, Andreas; Tegnered, Daniel; Nordman, Hans; Anderson, Johan; Strand, Pär

    2014-01-01

    Gyrokinetic simulations of ion temperature gradient mode and trapped electron mode driven impurity transport in a realistic tokamak geometry are presented and compared with results using simplified geometries. The gyrokinetic results, obtained with the GENE code in both linear and non-linear modes are compared with data and analysis for a dedicated impurity injection discharge at JET. The impact of several factors on heat and particle transport is discussed, lending special focus to tokamak geometry and rotational shear. To this end, results using s-alpha and concentric circular equilibria are compared with results with magnetic geometry from a JET experiment. To further approach experimental conditions, non-linear gyrokinetic simulations are performed with collisions and a carbon background included. The impurity peaking factors, computed by finding local density gradients corresponding to zero particle flux, are discussed. The impurity peaking factors are seen to be reduced by a factor of ~2 in realistic ge...

  17. Radial transport of toroidal angular momentum in tokamaks

    CERN Document Server

    Calvo, Ivan

    2014-01-01

    The radial flux of toroidal angular momentum is needed to determine tokamak intrinsic rotation profiles. Its computation requires knowledge of the gyrokinetic distribution functions and turbulent electrostatic potential to second-order in $\\epsilon = \\rho/L$, where $\\rho$ is the ion Larmor radius and $L$ is the variation length of the magnetic field. In this article, a complete set of equations to calculate the radial transport of toroidal angular momentum in any tokamak is presented. In particular, the $O(\\epsilon^2)$ equations for the turbulent components of the distribution functions and electrostatic potential are given for the first time without assuming that the poloidal magnetic field over the magnetic field strength is small.

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

    CERN Document Server

    Casati, Alessandro

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-01

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

  20. Design of geometric phase measurement in EAST Tokamak

    CERN Document Server

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

    2016-01-01

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

  1. The Aneutronic Rodless Ultra Low Aspect Ratio Tokamak

    Science.gov (United States)

    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.

  2. Imaging charge exchange recombination spectroscopy on the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

  3. Imaging charge exchange recombination spectroscopy on the TEXTOR tokamak

    Science.gov (United States)

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

    2010-12-01

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

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

    Science.gov (United States)

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

    2016-08-01

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

  5. Electromagnetic effects on rippling instability and tokamak edge fluctuations

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, Sadayoshi; Saleem, Hamid [National Inst. for Fusion Science, Toki, Gifu (Japan)

    1998-07-01

    Electromagnetic effects on rippling mode are investigated as a cause of low frequency electromagnetic fluctuations in tokamak edge region. It is shown that, in a current-carrying resistive plasma, the purely growing electrostatic rippling mode can turn out to be an electromagnetic oscillatory instability. The resistivity fluctuation and temperature gradient are the main sources of this instability, which requires both parallel and perpendicular wave vectors. The Alfven waves in a coupled dispersion relation are found heavily damped in such dissipative plasmas. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Castracane, J.

    2001-01-04

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

  7. HCN Laser Interferometer on the EAST Superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

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

    2008-01-01

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

  8. On Runaway Transport under Magnetic Turbulence in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Castejon, F.; Equilior, S.; Rodriguez-Rodrigo, L. [CIEMAT. Madrid (Spain)

    2001-07-01

    The influence of magnetic turbulence on runaway transport has been studied. The evolution of runaway distribution function has been calculated using Electra a 2D code in momentum space and 1D in radius coordinate. The code considers the effect of averaging the turbulence by runaway orbits. Then Hard X-Ray emission spectrum is estimated and compared with experimental results of TJ-1 tokamak, obtaining a remarkable agreement. (Author) 15 refs.

  9. Structural materials for large superconducting magnets for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Long, C.J.

    1976-12-01

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

  10. Application of advanced composites in tokamak magnet systems

    Energy Technology Data Exchange (ETDEWEB)

    Long, C. J.

    1977-11-01

    The use of advanced (high-modulus) composites in superconducting magnets for tokamak fusion reactors is discussed. The most prominent potential application is as the structure in the pulsed poloidal-field coil system, where a significant reduction in eddy currents could be achieved. Present low-temperature data on the advanced composites are reviewed briefly; they are too meager to do more than suggest a broad class of composites for a particular application.

  11. Analysis of neutral hydrogenic emission spectra in a tokamak

    Science.gov (United States)

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

    2015-10-01

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

  12. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    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. Stability-transport modeling of the SINP tokamak discharges

    Indian Academy of Sciences (India)

    S Lahiri; S Mukhopadhyay; A N S Iyengar; R Pal

    2001-05-01

    A one-dimensional stability transport code has been developed to simulate the evolution of tokamak plasma discharges. Explicit finite-difference methods have been used to follow the temporal evolution of the electron temperature equation. The poloidal field diffusion equation has been solved at every time step. The effects of MHD instabilities have been incorporated by solving equations for MHD mixing and tearing modes as and when required. The code has been applied to follow the evolution of tokamak plasma discharges obtained in the Saha Institute of Nuclear Physics (SINP) tokamak. From these simulations, we have been able to identify the possible models of thermal conductivity, diffusion and impurity contents in these discharges. Effects of different MHD modes have been estimated. It has been found that in low discharge =1, =1 and =2, =1 modes play major role in discharge evolution. These modes are found to result in the positive jump in the loop voltage which was also observed in the experiments. Hollow current density profile and negative shear in the profile have also been found in the rising phase of a discharge.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-06-01

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

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

    Science.gov (United States)

    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''.

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

    Science.gov (United States)

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

    2016-03-01

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

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

    Science.gov (United States)

    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.

  18. Optimization study of normal conductor tokamak for commercial neutron source

    Science.gov (United States)

    Fujita, T.; Sakai, R.; Okamoto, A.

    2017-05-01

    The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

  19. Numerical Tokamak Turbulence Calculations on the CRAY T3E

    Energy Technology Data Exchange (ETDEWEB)

    Lynch, V.E., Leboeuf, J.N., Carreras, B.A. [Oak Ridge National Lab., TN (United States)], Alvarez, J.D., Garcia, L. [Universidad `Carlos III` de Madrid (Spain)

    1997-12-31

    Full cross section calculations of ion-temperature-gradient-driven turbulence with Landau closure are being carried out as part of the Numerical Tokamak Turbulence Project, one of the U.S. Department of Energy`s Phase II Grand Challenges. To include the full cross section of a magnetic fusion device like the tokamak requires more memory and CPU time than is available on the National Energy Research Scientific Computing Center`s (NERSC`s) shared-memory vector machines such as the CRAY C90 and J90. Calculations of cylindrical multi-helicity ion-temperature-gradient-driven turbulence were completed on NERSC`s 160-processor distributed-memory CRAY T3E parallel computer with 256 Mbytes of memory per processor. This augurs well for yet more memory and CPU intensive calculations on the next-generation T3E at NERSC. This paper presents results on benchmarks with the current T3E at NERSC. Physics results pertaining to plasma confinement at the core of tokamaks subject to ion-temperature-gradient-driven-turbulence are also highlighted. Results at this resolution covering this extent of physical time were previously unattainable. Work is in progress to increase the resolution, improve the performance of the parallel code, and include toroidal geometry in these calculations in anticipation of the imminent arrival of a fully configured,512-processor, T3E-900 model.

  20. Modeling of Anomalous Transport in Tokamaks with FACETS code

    Science.gov (United States)

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

    2009-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Long-Poe Ku and Allen H. Boozer

    2009-06-05

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-17

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  4. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  5. Characterization of the Novillo Tokamak in main discharge regime; Caracterizacion del Tokamak Novillo en regimen de descarga principal

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-07-15

    The analytical procedure to carry out the establishment of the discharge in a Tokamak including: a) Ionization, b) Diffusion losses, recombination, union, drift speed, spurious fields, and c) Electric field is presented. In an experimental way a procedure settles down by means of which it is characterized the plasma, specially a new characteristic discharge parameter is settled down and it is the plasma current by the duration of the (I{sub p}t) discharge. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-15

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

  7. Computer Simulation of Transport Driven Current in Tokamaks

    Science.gov (United States)

    Nunan, William Joseph, III

    1995-01-01

    Plasma transport phenomena can drive large currents parallel to an externally applied magnetic field. The Bootstrap Current Theory accounts for the effect of Banana Diffusion on toroidal current, but the effect is not confined to that transport regime, or even to toroidal geometry. Our electromagnetic particle simulations have demonstrated that Maxwellian plasmas in static toroidal and vertical fields spontaneously develop significant toroidal current, even in the absence of the "seed current" which the Bootstrap Theory requires. Other simulations, in both cylindrical and toroidal geometries, and without any externally imposed electric field, show that if the plasma column is centrally fueled, then an initial toroidal current grows steadily, apparently due to a dynamo effect. The straight cylinder does not exhibit kink instabilities because k_ {z} = 0 in this 2 + 1/2 dimensional model. When the plasma is fueled at the edge rather than the center, the effect is diminished. Fueling at an intermediate radius should produce a level of current drive in between these two limits, because the key to the current drive seems to be the amount of total poloidal flux which the plasma crosses in the process of escaping. In a reactor, injected (cold) fuel ions must reach the center, and be heated up in order to burn; therefore, central fueling is needed anyway, and the resulting influx of cold plasma and outflux of hot plasma drives the toroidal current. Our simulations indicate that central fueling, coupled with the central heating due to fusion reactions may provide all of the required toroidal current. The Neoclassical Theory predicts that the Bootstrap Current approaches zero as the aspect ratio approaches infinity; however, in straight cylindrical plasma simulations, axial current increases over time at nearly the same rate as in the toroidal case. These results indicate that a centrally fueled and heated tokamak may sustain its own toroidal current, even in the absence of

  8. Including collisions in gyrokinetic tokamak and stellarator simulations

    Energy Technology Data Exchange (ETDEWEB)

    Kauffmann, Karla

    2012-04-10

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

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  10. Electron thermal transport in a scan of the effective ion charge in the RTP tokamak

    NARCIS (Netherlands)

    Konings, J. A.; Hogeweij, G. M. D.; Cardozo, N. J. L.; Oomens, A. A. M.; Schüller, F. C.

    1997-01-01

    A transport analysis is reported of a series of discharges in the Rijnhuizen Tokamak Project (RTP) tokamak, in which Z(eff) was varied by using mixtures of helium and neon as the filling gas. Z(eff) was scanned over the range 5.1 to 9.5. As a reference, pure deuterium plasmas with Z(eff) less than

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

    Institute of Scientific and Technical Information of China (English)

    吴斌; 张澄

    2003-01-01

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

  12. Systematic design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak

    NARCIS (Netherlands)

    Hennen, B.A.; Westerhof, E.; Nuij, Pwjm; M.R. de Baar,; Steinbuch, M.

    2012-01-01

    Suppression of tearing modes is essential for the operation of tokamaks. This paper describes the design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak. The two main control tasks of this feedback control system are the radial alignment of electron cyclot

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

    NARCIS (Netherlands)

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

    2008-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

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

    Directory of Open Access Journals (Sweden)

    E. Mazzucato

    2011-03-01

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

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

    Institute of Scientific and Technical Information of China (English)

    石秉仁

    2002-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    石秉仁

    2002-01-01

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

  18. Electron assisted glow discharges for conditioning fusion tokamak devices

    Science.gov (United States)

    Schaubel, K. M.; Jackson, G. L.

    1989-08-01

    Glow discharge conditioning of tokamaks with graphite plasma-facing surfaces has been used to reduce impurities and obtain density control of the plasma discharge. However, a major operational disadvantage of glow conditioning is the high pressure required to initiate the glow discharge, e.g., approx. 70 mTorr for helium in DIII-D, which requires isolating auxiliary components that can not tolerate the high pressure. An electron-gun assisted glow discharge can lower breakdown pressure, possibly eliminating the necessity of isolating these auxiliary systems during glow discharge conditioning and allowing glow discharge operation at lower pressures.

  19. Tokamak Transmutation of (nuclear) Waste (TTW): Parametric studies

    Science.gov (United States)

    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.

  20. Spectral measurements of runway electrons in the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kudyakov, Timur

    2009-07-22

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

  1. Halo current diagnostic system of experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Directory of Open Access Journals (Sweden)

    Vlainić Miloš

    2015-06-01

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

  3. Numerical simulation of internal reconnection event in spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takaya; Mizuguchi, Naoki; Sato, Tetsuya [National Inst. for Fusion Science, Toki, Gifu (Japan)

    1999-07-01

    Three-dimensional magnetohydrodynamic simulations are executed in a full toroidal geometry to clarify the physical mechanisms of the Internal Reconnection Event (IRE), which is observed in the spherical tokamak experiments. The simulation results reproduce several main properties of IRE. Comparison between the numerical results and experimental observation indicates fairly good agreements regarding nonlinear behavior, such as appearance of localized helical distortion, appearance of characteristic conical shape in the pressure profile during thermal quench, and subsequent appearance of the m=2/n=1 type helical distortion of the torus. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-14

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

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

    Science.gov (United States)

    Schaffer, Michael J.

    1986-01-01

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

  6. Halo current diagnostic system of experimental advanced superconducting tokamak

    Science.gov (United States)

    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.

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

    Institute of Scientific and Technical Information of China (English)

    罗家融

    2002-01-01

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

  8. Impact of poloidal convective cells on momentum flux in tokamaks

    Science.gov (United States)

    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.

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

    Institute of Scientific and Technical Information of China (English)

    郑永真

    2004-01-01

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

  10. Experimental measurement of electron heat diffusivity in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-06-01

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

  11. Controlling tokamak geometry with three-dimensional magnetic perturbations

    Energy Technology Data Exchange (ETDEWEB)

    Bird, T. M., E-mail: tbird@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, Wendelsteinstr. 1, 17491 Greifswald (Germany); Hegna, C. C. [Departments of Engineering Physics and Physics, University of Wisconsin-Madison, 1500 Engineering Dr., Madison, Wisconsin 53703 (United States)

    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.

  12. Negative edge plasma currents in the SINP tokamak

    Indian Academy of Sciences (India)

    Ramesh Narayanan; A N Sekar Iyengar

    2011-12-01

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

  13. Resonant magnetic perturbations and divertor footprints in poloidally diverted tokamaks

    CERN Document Server

    Cahyna, Pavel

    2010-01-01

    General formula describing both the divertor strike point splitting and width of magnetic islands created by resonant magnetic perturbations (RMPs) in a poloidally diverted tokamak equilibrium is derived. Under the assumption that the RMP is produced by coils at the low-field side such as those used to control edge localized modes (ELMs) it is demonstrated that the width of islands on different magnetic surfaces at the edge and the amount of divertor splitting are related to each other. Explanation is provided of aligned maxima of the perturbation spectra with the safety factor profile - an effect empirically observed in models of many perturbation coil designs.

  14. Improvement of tokamak confinement by current profile control

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, Kimitaka (National Inst. for Fusion Science, Nagoya (Japan)); Itoh, Sanae; Yagi, Masatoshi; Fukuyama, Atsushi; Azumi, Masafumi

    1993-12-01

    Impact of the current profile on the anomalous transport coefficients in tokamaks is discussed, based on the recent progress of the anomalous transport theory. When the central q-value is elevated above unity, the geometry turns to the magnetic well, and the anomalous transport is reduced. If the negative shear is realized, the anomalous transport is further reduced. The confinement improvement phenomena associated with the lower hybrid wave current drive and with high [beta][sub p] experiments are discussed as an application of this model. A motivation of the research on the steady state plasmas is also discussed. (author).

  15. Probability of statistical L-H transition in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, Sanae-I. [Kyushu Univ., Research Institute for Applied Mechanics, Kasuga, Fukuoka (Japan); Itoh, Kimitaka; Toda, Shinichiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2002-08-01

    A statistical model of bifurcation of radial electric field E{sub r} is analyzed in relation with L-H transitions of tokamaks. A noise from micro fluctuations leads to random noise for E{sub r}. The transition of E{sub r} occurs in a probabilistic manner. Probability density function and ensemble average of E{sub r} are obtained, when hysteresis of E{sub r} exists. Forward- and backward-transition probabilities are calculated. The phase boundary is shown. Due to the suppression of turbulence by E{sub r} shear, the boundary deviates from the Maxwell's construction rule. (author)

  16. Neutral Beam Injection Experiments in the HL-1M Tokamak

    Institute of Scientific and Technical Information of China (English)

    严龙文; 雷光玖; 钟光武; 江涛; 周艳; 姜韶风; 丁玄同; 周才品; 刘永

    2003-01-01

    Neutral beam injection (NBI) experiments have been carried out with two operation modes of a bucket ion source in the HL-1M tokamak. During the first mode, more than 30% rise in ion temperature above the Ohmic level is routinely achieved after NBI power about 0. 5 MW is injected. Ion temperature only increases 20-30% for the second operation mode, which is often limited by current termination. The heating effects of the NBI have been analysed experimentally and theoretically. The performance of the NBI system is well described.

  17. Gyrokinetic simulation of isotope scaling in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lee, W.W. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Santoro, R.A. [California Univ., Irvine, CA (United States). Dept. of Physics

    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. Lithium beam diagnostic system on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anda, G.; Bencze, A. [Wigner – RCP, HAS, Budapest (Hungary); Berta, M., E-mail: bertam@sze.hu [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Hacek, P. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Réfy, D.; Krizsanóczi, T.; Bató, S.; Ilkei, T.; Kiss, I.G.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

    2016-10-15

    Highlights: • Li-beam diagnostic system on the COMPASS tokamak is an improved and compact system to allow testing of Atomic Beam Probe. • The possibility to measure background corrected density profiles on the few microseconds time scale. • First Li-beam diagnostic system with recirculating neutralizer. • The system includes the redesigned ion source with longer lifetime. - Abstract: An improved lithium beam based beam emission spectroscopy system – installed on COMPASS tokamak – is described. The beam energy enhanced up to 120 keV for Atomic Beam Probe measurement. The size of the ion source is doubled, using a newly developed thermionic heater instead of the conventionally used heating (tungsten or molybdenum) filament. The neutralizer is also improved. It produces the same sodium vapor in a cell but minimize the loss condensing the vapor on a cold surface which is led back (in fluid state) into the sodium oven. This way we call it recirculating neutralizer. The observation system consists of a CCD camera and an avalanche photodiode array.

  19. Conceptual study of electron ripple injection for tokamak transport control

    Energy Technology Data Exchange (ETDEWEB)

    Choe, W.; Ono, M. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Chang, C.S. [New York Univ., NY (United States). Courant Inst. of Mathematical Sciences

    1995-08-01

    A non-intrusive method for inducing radial electric field based on electron ripple injection is under development by the Princeton CDX-U group. The radial electric field is known to play an important role in the L-H and H-VH mode transition according to the recent theoretical and experimental research. It is therefore important to develop a non-intrusive tool to control the radial electric field 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 {gradient}B and curvature drifts, causing the flux surfaces to charge up negatively. Electron cyclotron resonance heating is utilized to increase the trapped population and the electron drift velocity by raising the perpendicular energy of trapped electrons. In order to quantify the effects of cyclotron resonance heating on electrons, the temperature anisotropy of resonant electrons in a tokamak plasma is calculated. For the calculation of anisotropic temperatures, energy moments of the bounce-averaged Fokker-Planck equation with a bi-Maxwellian distribution function for heated electrons are solved, assuming a moderate wave power and a constant quasilinear diffusion coefficient. Simulation 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 parameters are given.

  20. The timing system on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

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

  1. Analysis of fast ion induced instabilities in tokamak plasmas

    CERN Document Server

    Horváth, László

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Klyuchnikov, L. A., E-mail: lklyuchnikov@list.ru; Krupin, V. A.; Nurgaliev, M. R.; Korobov, K. V.; Nemets, A. R.; Dnestrovskij, A. Yu.; Tugarinov, S. N.; Serov, S. V. [National Research Centre “Kurchatov Institute,” Moscow (Russian Federation); Naumenko, N. N. [B.I. Stepanov Institute of Physics NASB, Minsk, Republic of Belarus (Belarus)

    2016-05-15

    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 H{sub α} beam line. Ion temperature measurements are routinely provided using the C{sup 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.

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

    Science.gov (United States)

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

    2016-05-01

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

  4. The vacuum vessel thermal shield of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-09-01

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

  5. Overview of the Pegasus Extremely Low-Aspect Ratio Tokamak

    Science.gov (United States)

    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 ≈ 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.

  6. Gyrokinetic theory and dynamics of the tokamak edge

    Energy Technology Data Exchange (ETDEWEB)

    Scott, B. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2016-08-15

    The validity of modern gyrokinetic field theory is assessed for the tokamak edge. The basic structure of the Lagrangian and resulting equations and their conservation laws is reviewed. The conventional microturbulence ordering for expansion is small potential/arbitrary wavelength. The equilibrium ordering for expansion is long wavelength/arbitrary amplitude. The long-wavelength form of the conventional Lagrangian is derived in detail. The two Lagrangians are shown to match at long wavelength if the E x B Mach number is small enough for its corrections to the gyroaveraging to be neglected. Therefore, the conventional derivation and its Lagrangian can be used at all wavelengths if these conditions are satisfied. Additionally, dynamical compressibility of the magnetic field can be neglected if the plasma beta is small. This allows general use of a shear-Alfven Lagrangian for edge turbulence and self consistent equilibrium-scale phenomena for flows, currents, and heat fluxes for conventional tokamaks without further modification by higher-order terms. Corrections in polarisation and toroidal angular momentum transport due to these higher-order terms for global edge turbulence computations are shown to be small. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. Localised measurements of turbulence in the Tore Supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Devynck, P.; Garbet, X.; Laviron, C.; Payan, J.; Saha, S.K. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Gervais, F.; Hennequin, P.; Quemeneur, A.; Truc, A. [Ecole Polytechnique, 91 - Palaiseau (France). Lab. de Physique des Milieux Ionises

    1992-04-01

    A collective infra-red laser scattering diagnostic has been installed on the TORE SUPRA tokamak for the measurement of plasma density fluctuations. For the range of wave numbers explored (3-15 cm{sup -1}), the scattering angles are very weak (about 1 mrad). Consequently, the scattering signals are averaged along the whole observation chord, resulting in poor longitudinal spatial localisation. However, by virtue of the pitch angle variation of the magnetic field lines in the tokamak, and of the perpendicularity of the turbulence wave vector to these field lines, it has been possible to obtain partial spatial resolution along the direction of the beam. Good agreement between the experimental and theoretical angular resolution of the diagnostic as well as the results of cross-correlation performed on the signals obtained by two simultaneous probing beams also justify this novel concept. From the variation of the fluctuation power with the orientation angle of the observed wave sector, it has been possible to deduce the radial fluctuation profile by a deconvolution procedure, showing that the fluctuations increase sharply near the edge. The k-spectrum was also measured and shows a k{sup -3} dependence for k>6 cm{sup -1}. Experimental evidences are put forward to show that the k-spectrum is neither purely poloidal nor purely radial in the (k{sub r}-k{sub {theta}}) plane.

  8. Electron ripple injection concept for tokamak transport control

    Science.gov (United States)

    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.

  9. Resistive Edge Modes in Stellarator and Tokamak Geometries

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

    The reactive ion-temperature-gradient driven drift mode (or mode) is a promising candidate for explaining the anomalous transport in the core of tokamak plasmas. However, a strong influence of electron-ion collisions in the edge region gives a resistive nature to the drift modes. So far, a lot of work has been done towards understanding of these modes in tokamak configurations, whereas a limited amount of work has been reported in stellarators. In the present work, linear stability of the collisional mode and the resistive ballooning mode in the electrostatic limit is studied in a three-dimensional Wendelstein 7-X Stellarator geometry. The full magnetic field configuration is obtained using the variational moments equilibrium code VMEC. The reduced Braghinskii equations are used as a model for the electrons and an advanced fluid model for the ions. By employing the ballooning mode formalism, the drift wave problem is set as an eigenvalue equation along a field line. The derived eigenvalue equation is solved numerically using a standard shooting technique and applying WKB type boundary conditions. The growth rates and real frequencies of the most unstable modes and their eigenfunctions are calculated. The effects of collisions, density and temperature gradients and other geometrical quantities on mode localization and stability are studied. Finally, the results are contrasted and compared with those obtained for an ITER-like geometry. (Author)

  10. Steady-state operation in compact tokamaks with copper coils

    Science.gov (United States)

    Kuteev, B. V.; Azizov, E. A.; Bykov, A. S.; Dnestrovsky, A. Yu.; Dokuka, V. N.; Gladush, G. G.; Golikov, A. A.; Goncharov, P. R.; Gryaznevich, M.; Gurevich, M. I.; Ivanov, A. A.; Khairutdinov, R. R.; Khripunov, V. I.; Kingham, D.; Klishchenko, A. V.; Kurnaev, V. A.; Lukash, V. E.; Medvedev, S. Yu.; Savrukhin, P. V.; Sergeev, V. Yu.; Shpansky, Yu. S.; Sykes, A.; Voss, G.; Zhirkin, A. V.

    2011-07-01

    This paper considers a fast track to non-energy applications of nuclear fusion that is associated with the 'fusion for neutrons' (F4N) paradigm. Being a useful product accompanying energy, fusion neutrons are more valuable than the energy released in DT reactions and they are urgently needed for research purposes and to develop and validate modern technologies. In the near future neutron yield in fusion devices will become significantly larger than that of fission and accelerator sources. This paper describes a compact tokamak fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ~0.5 and ~0.3 m with magnetic field ~1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The production rate of DT neutrons of (3-10) × 1017 n s-1 and their flux at the first wall of 0.2 MW m-2 ensure that the device is capable of fusion-fission demonstration experiments. The problems of major concern are discharge initiation, current drive, plasma—fast ion beam stability and high first wall and divertor loads. The conceptual design provides solutions to these problems and suggests the feasibility of the FNS-ST.

  11. Measurement of the effective plasma ion mass in large tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Lister, J.B.; Villard, L.; Ridder, G. de [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP)

    1997-06-01

    There is not yet a straightforward method for the measurement of the D-T ratio in the centre of a tokamak plasma. One of the simpler measurements put forward in the past is the interpretation of the MHD spectrum in the frequency range of the Global Alfven Eigenmodes (GAE). However, the frequencies of these modes do not only depend on the plasma mass, but are also quite strongly dependent on the details of the current and density profiles, creating a problem of deconvolution of the estimate of the plasma mass from an implicit relationship between several measurable plasma parameters and the detected eigenmode frequencies. This method has been revised to assess its likely precision for the JET tokamak. The low n GAE modes are sometimes too close to the continuum edge to be detectable and the interpretation of the GAE spectrum is rendered less direct than had been hoped. We present a statistical study on the precision with which the D-T ratio could be estimated from the GAE spectrum on JET. (author) 4 figs., 8 refs.

  12. ECE RADIOMETER UPGRADE ON THE DIII-D TOKAMAK

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Plasma engineering studies for Tennessee Tokamak (TENTOK) fusion power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, K.E.; Lacatski, J.T.; Miller, J.B.; Bryan, W.E.; King, P.W.; Santoro, R.T.; Uckan, N.A.; Shannon, T.E.

    1984-02-01

    This paper summarizes the results of the plasma engineering and systems analysis studies for the Tennessee Tokamak (TENTOK) fusion power reactor. TENTOK is a 3000-MW(t) central station power plant that uses deuterium-tritium fuel in a D-shaped tokamak plasma configuration with a double-null poloidal divertor. The major parameters are R/sub 0/ = 6.4 m, a = 1.6 m, sigma (elongation) = 1.65, (n) = 1.5 x 10/sup 20/ m/sup -3/, (T) = 15 keV, (..beta..) = 6%, B/sub T/ (on-axis) = 5.6 T, I/sub p/ = 8.5 MA, and wall loading = 3 MW/m/sup 2/. Detailed analyses are performed in the areas of (1) transport simulation using the one-and-one-half-dimensional (1-1/2-D) WHIST transport code, (2) equilibrium/poloidal field coil systems, (3) neutral beam and radiofrequency (rf) heating, and (4) pellet fueling. In addition, impurity control systems, diagnostics and controls, and possible microwave plasma preheating and steady-state current drive options are also considered. Some of the major features of TENTOK include rf heating in the ion cyclotron range of frequencies, superconducting equilibrium field coils outside the superconducting toroidal field coils, a double-null poloidal divertor for impurity control and alpha ash removal, and rf-assisted plasma preheating and current startup.

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

    Science.gov (United States)

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

    2016-06-01

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

  15. A Research Program of Spherical Tokamak in China

    Institute of Scientific and Technical Information of China (English)

    何也熙

    2002-01-01

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

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

    Science.gov (United States)

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

    2015-11-01

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

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

    Science.gov (United States)

    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.

  18. Active cooling system for Tokamak in-vessel operation manipulator

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang

    2015-10-15

    Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.

  19. Dynamic simulations of the cryogenic system of a tokamak

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Halpern, F.D., E-mail: federico.halpern@epfl.ch [Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne (Switzerland); Ricci, P.; Jolliet, S. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne (Switzerland); Loizu, J. [Max-Planck-Institut für Plasmaphysik, D-17491, Greifswald (Germany); Morales, J.; Mosetto, A.; Musil, F.; Riva, F.; Tran, T.M.; Wersal, C. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne (Switzerland)

    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.

  1. Nuclear shielding of openings in ITER Tokamak building

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  4. Pellet injection and confinement in the tore supra tokamak; Injection de glacons et confinement dans le tokamak tore supra

    Energy Technology Data Exchange (ETDEWEB)

    Maget, P

    1998-09-23

    Pellet injection in the centre of tokamak plasmas can lead to an improved confinement regime called PEP (Pellet Enhanced Performance). The present work is dedicated to the mechanisms involved in the PEP regimes obtained in the tokamak Tore Supra. A neoclassical approach of transport shows that it is the anomalous transport, due to plasma turbulence, that causes the enhanced confinement. A linear model describing electrostatic instabilities has been developed in order to study the roles of density profile and current profile during the PEP, in the limit of large growth rates. The effect ofradial shear in flows is taken into account by removing the ExB shear flow rate from the linear growth rate, as suggested by non-linear numerical simulations of turbulence. A local transport coefficient is estimated from the knowledge of the linear growth rate and the mode width. We find that the peaked density profile in PEP regime lowers the diffusion coefficient, and that the velocity shear amplifies this effect. The evolution of the current profile is also stabilizing, but this parameter is not known with sufficient accuracy, so that its role in Tore Supra PEP experiments remains uncertain. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

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

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

    CERN Document Server

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Peters, M.

    1996-01-16

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

  8. Overview of the TCV tokamak program: scientific progress and facility upgrades

    DEFF Research Database (Denmark)

    Coda, S.; Ahn, J.; Albanese, R.

    2017-01-01

    The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range with...

  9. Overview of the ITER Tokamak complex building and integration of plant systems toward construction

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2015-10-15

    The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed

  10. A relativistic model of electron cyclotron current drive efficiency in tokamak plasmas

    OpenAIRE

    Lin-Liu Y.R.; Hu Y.J.; Hu Y.M.

    2012-01-01

    A fully relativistic model of electron cyclotron current drive (ECCD) efficiency based on the adjoint function techniques is considered. Numerical calculations of the current drive efficiency in a tokamak by using the variational approach are performed. A fully relativistic extension of the variational principle with the modified basis functions for the Spitzer function with momentum conservation in the electron-electron collision is described in general tokamak geometry. The model developed ...

  11. Development of the Fast Ionization Gauge in the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    WANGMingxu; LIBo; YANGZhigang; LIAOZhiqing; YANLongwen; ZHANGNianman; YANDonghai

    2003-01-01

    The neutral gas pressure near plasma or divertor plates is very important for the plasma-wall interaction, which determine the operation mode of divertom and confinement performances of plasma in tokamaks. The commercial ionization gauge does not work in strong magnetic field and noisy enviroment encountered in tokamaks. The measuring errom of pressure commercial ionizationare very large by the gauge mounted on the pumping system or through a long pipe to the vacuum vessel. A new ionization gauge,

  12. Potential Safe Termination by Laser Ablation of High Z Impurity in the HL-1M Tokamak

    Institute of Scientific and Technical Information of China (English)

    ZHENGYongzhen; FENGXingya; ZHENGYinjja; GUOGancheng; XUDeming; DENGZhongchao

    2003-01-01

    In the contemporary large tokamak, the disruptive termination of a discharge will reduce the lifetime of the first wall materials with the intense heat flux at the energy quench and the intense runaway electrons duringthe current quench, and generate high electron magnetic forces on vacuum vessel components with intense eddy current at the current quench. Thus, avoidance and softening of the energy quench and the current quench and controlling an expected disruption or emergency shutdown must be established in the present tokamak machines.

  13. HL-2A tokamak disruption forecasting based on an artificial neural network

    Institute of Scientific and Technical Information of China (English)

    Wang Hao; Wang Ai-Ke; Yang Qing-Wei; Ding Xuan-Tong; Dong Jia-Qi; Sanuki H; Itoh K

    2007-01-01

    Artificial neural networks are trained to forecast the plasma disruption in HL-2A tokamak. Optimized network architecture is obtained. Saliency analysis is made to assess the relative importance of different diagnostic signals as network input. The trained networks can successfully detect the disruptive pulses of HL-2A tokamak. The results obtained show the possibiliry of developing a neural network predictor that intervenes well in edvance for avoiding plasma disruption or mitigating its effects.

  14. Alternate Data Acquisition and Real-time Monitoring System on HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    Wei Peijie; Luo Jiarong; Wang Hua; Li Guiming

    2005-01-01

    A new system called alternate data acquisition and real-time monitoring system has been developed for long-time discharge in tokamak operation. It can support continuous on-line data acquisition at a high sampling rate and a graphic display of the plasma parameters during the discharge. Thus operators can monitor and control the plasma state in real time. An application of this system has been demonstrated on the HT-7 tokamak.

  15. Axisymmetric instability in a noncircular tokamak: experiment and theory

    Energy Technology Data Exchange (ETDEWEB)

    Lipschultz, B.; Prager, S.C.; Todd, A.M.M.; Delucia, J.

    1979-09-01

    The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes -- the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria. Experimentally, the square is vertically stable and both dee's unstable to a vertical nonrigid axisymmetric shift. The central magnetic axis displacement grows exponentially with a growth time approximately 10/sup 3/ poloidal Alfven times plasma time. Proper initial positioning of the plasma on the midplane allows passive feedback to nonlinearly restore vertical motion to a small stable oscillation. Experimental poloidal flux plots are produced directly from internal magnetic probe measurements.

  16. Small angle slot divertor concept for long pulse advanced tokamaks

    Science.gov (United States)

    Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

    2017-04-01

    SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

  17. WILDCAT: a catalyzed D-D tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

    1981-11-01

    WILDCAT is a conceptual design of a catalyzed D-D, tokamak, commercial, fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing D-T designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete, conceptual design.

  18. A divertor plasma configuration design method for tokamaks

    Science.gov (United States)

    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).

  19. GPEC, a real-time capable Tokamak equilibrium code

    CERN Document Server

    Rampp, Markus; Fischer, Rainer

    2015-01-01

    A new parallel equilibrium reconstruction code for tokamak plasmas is presented. GPEC allows to compute equilibrium flux distributions sufficiently accurate to derive parameters for plasma control within 1 ms of runtime which enables real-time applications at the ASDEX Upgrade experiment (AUG) and other machines with a control cycle of at least this size. The underlying algorithms are based on the well-established offline-analysis code CLISTE, following the classical concept of iteratively solving the Grad-Shafranov equation and feeding in diagnostic signals from the experiment. The new code adopts a hybrid parallelization scheme for computing the equilibrium flux distribution and extends the fast, shared-memory-parallel Poisson solver which we have described previously by a distributed computation of the individual Poisson problems corresponding to different basis functions. The code is based entirely on open-source software components and runs on standard server hardware and software environments. The real-...

  20. Internal transport barrier formation in the Tore Supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Turco, F; Giruzzi, G; Artaud, J F; Mazon, D [Association Euratom-CEA sur la Fusion, CEA/DSM/IRFM, Cadarache, 13108 St. Paul-lez-Durance (France)

    2009-06-15

    The characteristics of the safety factor profile (q) have been studied for a database of discharges featuring internal transport barrier phenomena under different heating and current drive conditions in the Tore Supra tokamak. The presence of a recurrent link between the time of formation of an internal barrier and the central safety factor (q{sub 0}) crossing a low order rational surface is reported on and discussed for a database of {approx}40 discharges. Other relevant features of the current profile (e.g. the minimal q value, the magnetic shear and the presence of magnetic islands in the plasma) have been investigated, yielding to little or no correlation with the time and location of the barrier.

  1. Blobs in the tokamak scrape-off layer

    Science.gov (United States)

    Jovanovic, D.; Shukla, P. K.; Pegoraro, F.

    2008-07-01

    A three-dimensional model for the warm-ion turbulence at the tokamak edge plasma and in the scrape-off layer is proposed. It is based on the nonlinear interchange mode, coupled with the nonlinear resistive drift mode, in the presence of the magnetic curvature drive, the density inhomogeneity, the electron dynamics along the open magnetic field lines, and the electron-ion and electron-neutral collisions. Numerical solutions indicate the collapse of the blob in the lateral direction, followed by a clockwise rotation and radial propagation. The symmetry breaking, caused both by the parallel resistivity and the finite ion temperature, introduces a poloidal component in the plasma blob propagation, while the overall stability properties and the speed are not affected qualitatively.

  2. Remote network control plasma diagnostic system for Tokamak T-10

    Science.gov (United States)

    Troynov, V. I.; Zimin, A. M.; Krupin, V. A.; Notkin, G. E.; Nurgaliev, M. R.

    2016-09-01

    The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet.

  3. Engineering aspects of the HT-6M Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    1986-05-01

    The HT-6M is a medium-sized tokamak being built in China. The principal aim of the project is to study high-power auxiliary heating (1-MW neutral beam injection, 1-MW ion cyclotron resonance heating, and 100-kW electron cyclotron resonance heating), high-..beta.. experiments, the transport process, and the formation and diffusion process of impurities. The main device parameters are: major plasma radius R = 65 cm, minor plasma radius a = 20 cm, plasma current I/subP/ = 150 kA, discharge time tau = 150 ms, toroidal field B/subT/ = 15 kG. Simplicity of construction, accessibility to the plasma, reliability in operation, and convenience for maintenance were particularly emphasized in the design. The important design features of the device and power supply system are described.

  4. Development and Validation of a Tokamak Skin Effect Transformer model

    CERN Document Server

    Romero, J A; Coda, S; Felici, F; Garrido, I

    2012-01-01

    A control oriented, lumped parameter model for the tokamak transformer including the slow flux penetration in the plasma (skin effect transformer model) is presented. The model does not require detailed or explicit information about plasma profiles or geometry. Instead, this information is lumped in system variables, parameters and inputs. The model has an exact mathematical structure built from energy and flux conservation theorems, predicting the evolution and non linear interaction of the plasma current and internal inductance as functions of the primary coil currents, plasma resistance, non-inductive current drive and the loop voltage at a specific location inside the plasma (equilibrium loop voltage). Loop voltage profile in the plasma is substituted by a three-point discretization, and ordinary differential equations are used to predict the equilibrium loop voltage as function of the boundary and resistive loop voltages. This provides a model for equilibrium loop voltage evolution, which is reminiscent ...

  5. Stabilization of the resistive shell mode in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Fitzpatrick, R.; Aydemir, A.

    1995-02-01

    The stability of current-driven external-kink modes is investigated in a tokamak plasma surrounded by an external shell of finite electrical conductivity. According to conventional theory, the ideal mode can be stabilized by placing the shell sufficiently close to the plasma, but the non-rotating ``resistive shell mode,`` which grows on the characteristic L/R time of the shell, always persists. It is demonstrated, using both analytic and numerical techniques, that a combination of strong edge plasma rotation and dissipation somewhere inside the plasma is capable of stabilizing the resistive shell mode. This stabilization mechanism does not necessarily depend on toroidicity or presence of resonant surfaces inside the plasma.

  6. Transport Bifurcation Induced by Sheared Toroidal Flow in Tokamak Plasmas

    CERN Document Server

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

    2011-01-01

    First-principles numerical simulations are used to describe a transport bifurcation in a differentially rotating tokamak plasma. Such a bifurcation is more probable in a region of zero magnetic shear, where the component of the sheared toroidal flow that is perpendicular to the magnetic field has the strongest suppressing effect on the turbulence, than one of finite magnetic shear. Where the magnetic shear is zero, there are no growing linear eigenmodes at any finite value of flow shear. However, subcritical turbulence can be sustained, owing to the transient growth of modes driven by the ion temperature gradient (ITG) and the parallel velocity gradient (PVG). Nonetheless, in a parameter space containing a wide range of temperature gradients and velocity shears, there is a sizeable window where all turbulence is suppressed. Combined with the relatively low transport of momentum by collisional (neoclassical) mechanisms, this produces the conditions for a bifurcation from low to high temperature and velocity gr...

  7. Gyrokinetic theory for arbitrary wavelength electromagnetic modes in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Qin, H.; Tang, W.M.; Rewoldt, G.

    1997-10-15

    A linear gyrokinetic system for arbitrary wavelength electromagnetic modes is developed. A wide range of modes in inhomogeneous plasmas, such as the internal kink modes, the toroidal Alfven eigenmode (TAE) modes, and the drift modes, can be recovered from this system. The inclusion of most of the interesting physical factors into a single framework enables one to look at many familiar modes simultaneously and thus to study the modifications of and the interactions between them in a systematic way. Especially, the authors are able to investigate self-consistently the kinetic MHD phenomena entirely from the kinetic side. Phase space Lagrangian Lie perturbation methods and a newly developed computer algebra package for vector analysis in general coordinate system are utilized in the analytical derivation. In tokamak geometries, a 2D finite element code has been developed and tested. In this paper, they present the basic theoretical formalism and some of the preliminary results.

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

    CERN Document Server

    Kuiroukidis, Ap; Tasso, H

    2015-01-01

    By choosing appropriate deformed Maxwellian ion and electron distribution functions depending on the two particle constants of motion, i.e. the energy and toroidal angular momentum, we reduce the Vlasov axisymmetric equilibrium problem for quasineutral plasmas to a transcendental Grad-Shafranov-like equation. This equation is then solved numerically under the Dirichlet boundary condition for an analytically prescribed boundary possessing a lower X-point to construct tokamak equilibria with toroidal sheared ion flow and anisotropic pressure. Depending on the deformation of the distribution functions these steady states can have toroidal current densities either peaked on the magnetic axis or hollow. These two kinds of equilibria may be regarded as a bifurcation in connection with symmetry properties of the distribution functions on the magnetic axis.

  9. Investigations of low discharges in the SINP tokamak

    Indian Academy of Sciences (India)

    S Lahiri; A N S Iyengar; S Kukhopadhyay; R Pal

    2002-01-01

    Low edge safety factor discharges including very low (1 < < 2) and ultra low (0 < < 1) have been obtained in the SINP tokamak. It has been observed that accessibility of these discharges depends crucially on the fast rate of plasma current rise. Several interesting results in terms of different time scales like , etc have been obtained using a set of softwares developed at SINP. From fluctuation analysis of the external magnetic probe data it has been found that MHD instabilities = 1, = 1 and = 2, = 1 etc. play major role in the evolution of these discharges. To investigate the internal details of these discharges, an internal magnetic probe system has been developed using which current density and other related parameters have been estimated. By carrying out a resistive stability analysis, evidence of the above-mentioned MHD instabilities have again been found. The physical processes lying behind the accessibility and evolution of the low discharges have been thoroughly investigated.

  10. Improved Mirnov Magnetic Coils System for the TCABR Tokamak

    Science.gov (United States)

    Vannucci, Alvaro; Olschewski, Erich; Kuznetsov, Yuri; Kucinski, Mutsuko; Tadeu Degasperi, Francisco; Araujo, Mauro Sergio; Galvao, Ricardo; Okano, Valdir; Nascimento, Ivan

    2000-10-01

    The Mirnov magnetic coils system for the TCABR was recently reconstructed. The most interesting aspect of this system is that the measured experimental signals already incorporate the influence of the toroidal geometry. This means that the usual fast Fourier transform analysis done on the magnetic experimental data is able to indicate, more precisely and in a straightforward way, the MHD mode contribution to the detected signals during a plasma discharge. The influence of the toroidal geometry on the Fourier analysis of the magnetic signals was investigated by carring a series of simulations, considering the Merezhkin correction expressed only as a function of the inverse of the tokamak aspect ratio (calculated at the position of interest). The results obtained clearly showed the existence of a phase modulation on the Mirnov signals which is not usually considered when the magnetic signals are Fourier analyzed in the frame of cylindrical approximation, that is, by neglecting the existing toroidal effect.

  11. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

  12. Systems study of tokamak fusion--fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations.

  13. Micro-tearing modes in the Mega Ampere Spherical Tokamak

    CERN Document Server

    Applegate, D J; Connor, J W; Cowley, S C; Dorland, W; Hastie, R J; Joiner, N; 10.1088/0741-3335/49/8/001

    2011-01-01

    Recent gyrokinetic stability calculations have revealed that the spherical tokamak is susceptible to tearing parity instabilities with length scales of a few ion Larmor radii perpendicular to the magnetic field lines. Here we investigate this 'micro-tearing' mode in greater detail to uncover its key characteristics, and compare it with existing theoretical models of the phenomenon. This has been accomplished using a full numerical solution of the linear gyrokinetic-Maxwell equations. Importantly, the instability is found to be driven by the free energy in the electron temperature gradient as described in the literature. However, our calculations suggest it is not substantially affected by either of the destabilising mechanisms proposed in previous theoretical models. Instead the instability is destabilised by interactions with magnetic drifts, and the electrostatic potential. Further calculations reveal that the mode is not significantly destabilised by the flux surface shaping or the large trapped particle f...

  14. How to upgrade a control system for a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Tenten, W. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany)); Dohms, U. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany)); Fuss, L. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany)); Huppertz, H. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany)); Lerch, J. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany)); Mueller, K.D. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany)); Reinhart, P. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany)); Rongen, F. (Zentrallabor fuer Elektronik, Forschungszentrum Juelich GmbH (KFA), 52425 Juelich (Germany))

    1994-12-15

    The TEXTOR tokamak for technology-oriented research in Juelich has been in operation since 1981. Its control system consists basically of a CAMAC computer system (PDP-11) for remote control and display, linked to programmable controllers (SIEMENS S3) for subsystem control via fibre optic cables. Due to several reasons, an upgrade of this well-established control system has become unavoidable. The main objective for this process is to provide better availability and reliability for another decade of operation and to reduce maintenance costs significantly. In this respect all CAMAC instrumentation had to be preserved completely. The paper describes in detail the background, design and layout of the new control system. Because upgrading an existing control system substantially differs from constructing a new system for a new device, special attention is given to the steps of achieving a smooth upgrade procedure that avoids unnecessary interferences with the TEXTOR operation. ((orig.))

  15. Turbulent transport of alpha particles in tokamak plasmas

    Science.gov (United States)

    Croitoru, A.; Palade, D. I.; Vlad, M.; Spineanu, F.

    2017-03-01

    We investigate the \\boldsymbol{E}× \\boldsymbol{B} diffusion of fusion born α particles in tokamak plasmas. We determine the transport regimes for a realistic model that has the characteristics of the ion temperature gradient (ITG) or of the trapped electron mode (TEM) driven turbulence. It includes a spectrum of potential fluctuations that is modeled using the results of the numerical simulations, the drift of the potential with the effective diamagnetic velocity and the parallel motion. Our semi-analytical statistical approach is based on the decorrelation trajectory method (DTM), which is adapted to the gyrokinetic approximation. We obtain the transport coefficients as a function of the parameters of the turbulence and of the energy of the α particles. According to our results, significant turbulent transport of the α particles can appear only at energies of the order of 100 KeV. We determine the corresponding conditions.

  16. Gyrokinetic modelling of stationary electron and impurity profiles in tokamaks

    CERN Document Server

    Skyman, Andreas; Tegnered, Daniel

    2014-01-01

    Particle transport due to Ion Temperature Gradient/Trapped Electron (ITG/TE) mode turbulence is investigated using the gyrokinetic code GENE. Both a reduced quasilinear (QL) treatment and nonlinear (NL) simulations are performed for typical tokamak parameters corresponding to ITG dominated turbulence. A selfconsistent treatment is used, where the stationary local profiles are calculated corresponding to zero particle flux simultaneously for electrons and trace impurities. The scaling of the stationary profiles with magnetic shear, safety factor, electron-to-ion temperature ratio, collisionality, toroidal sheared rotation, triangularity, and elongation is investigated. In addition, the effect of different main ion mass on the zero flux condition is discussed. The electron density gradient can significantly affect the stationary impurity profile scaling. It is therefore expected, that a selfconsistent treatment will yield results more comparable to experimental results for parameter scans where the stationary b...

  17. RF wave propagation and scattering in turbulent tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Horton, W., E-mail: wendell.horton@gmail.com; Michoski, C. [Institute for Fusion Studies, The University of Texas at Austin, Austin, TX 78654 (United States); Peysson, Y.; Decker, J. [CEA, IRFM, 13108, Saint-Paul, Durance Cedex (France)

    2015-12-10

    Drift wave turbulence driven by the steep electron and ion temperature gradients in H-mode divertor tokamaks produce scattering of the RF waves used for heating and current drive. The X-ray emission spectra produced by the fast electrons require the turbulence broaden RF wave spectrum. Both the 5 GHz Lower Hybrid waves and the 170 GHz electron cyclotron [EC] RF waves experience scattering and diffraction by the electron density fluctuations. With strong LHCD there are bifurcations in the coupled turbulent transport dynamics giving improved steady-state confinement states. The stochastic scattering of the RF rays makes the prediction of the distribution of the rays and the associated particle heating a statistical problem. Thus, we introduce a Fokker-Planck equation for the probably density of the RF rays. The general frame work of the coupled system of coupled high frequency current driving rays with the low-frequency turbulent transport determines the profiles of the plasma density and temperatures.

  18. Tokamak power reactor ignition and time dependent fractional power operation

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.

  19. Results from deuterium-tritium tokamak confinement experiments

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.

    1997-02-01

    Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which enabled the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation which have both good confinement and are MHD stable have enabled a broad study of burning plasma physics issues. A review of the technical and scientific results from the deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) is given with particular emphasis on alpha-particle physics issues.

  20. Magnetic Fluctuation Measurement in Sino United Spherical Tokamak Plasma

    Institute of Scientific and Technical Information of China (English)

    LIU Fei; WANG Wen-Hao; HE Ye-Xi; LIU Jun; TAN Yi; XIE Li-Feng; ZENG Long

    2007-01-01

    To investigate the magnetic fluctuations and for further transport study, the poloidal and radial magnetic field measurement is conducted on the Sino United Spherical Tokamak (SUNIST). Auto-power spectral density indicates that the magnetic fluctuation energy mainly concentrates in the frequency region lower than 10kHz. The magnetic field oscillations, which are characterized by harmonic frequencies of 40 kHz, are observed in the scrapeoff layer; by contrast, in the plasma core, the magnetic fluctuations are of Gaussian type. The time-frequency profiles show that the poloidal magnetic fluctuations are temporally intermittent. The autocorrelation calculation indicates that the fluctuations in decorrelation time vary between the core and the edge.

  1. Theory of self-organized critical transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Kishimoto, Y.; Tajima, T.; Horton, W.; LeBrun, M.J.; Kim, J.Y. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment]|[Texas Univ., Austin, TX (United States). Inst. for Fusion Studies

    1995-07-01

    A theoretical and computational study of the ion temperature gradient and {eta}{sub i} instabilities in tokamak plasmas has been carried out. In toroidal geometry the modes have a radially extended structure and their eigenfrequencies are constant over many rational surfaces that are coupled through toroidicity. These nonlocal properties of the ITG modes impose strong constraint on the drift mode fluctuations and the amciated transport, showing a self-organized characteristic. As any significant deviation away from marginal stability causes rapid temperature relaxation and intermittent bursts, the modes hover near marginality and exhibit strong kinetic characteristics. As a result, the temperature relaxation is self-semilar and nonlocal, leading to a radially increasing heat diffusivity. The nonlocal transport leads to the Bohm-like diffusion scaling. The heat input regulates the deviation of the temperature gradient away from marginality. The obtained transport scalings and properties are globally consistent with experimental observations of L-mode charges.

  2. Deposition of fuel pellets injected into tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Baylor, L.R.; Jernigan, T.C. [Oak Ridge National Lab., TN (United States); Hsieh, C. [General Atomics, San Diego, CA (United States)

    1998-06-01

    Pellet injection has been used on tokamak devices in a number of experiments to provide plasma fueling and density profile control. The mass deposition of these fuel pellets defined as the change in density profile caused by the pellet, has been found to show an outward displacement of the ablated material from that expected by mapping the theoretical ablation rate onto the flux surfaces. This suggests that fast transport of the pellet ablatant occurs during the flow along field lines that may be driven by {del}B drift effects. A comparison of the deposition of pellets from different machines shows similar behavior. Initial results from alternative injection locations designed to take advantage of the outward ablatant drift is presented.

  3. Tokamak fusion reactors with less than full tritium breeding

    Energy Technology Data Exchange (ETDEWEB)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed.

  4. Effect of density changes on tokamak plasma confinement

    CERN Document Server

    Spineanu, F

    2015-01-01

    A change of the particle density (by gas puff, pellets or impurity seeding) during the plasma discharge in tokamak produces a radial current and implicitly a torque and rotation that can modify the state of confinement. After ionization the newly born ions will evolve toward the periodic neoclassical orbits (trapped or circulating) but the first part of their excursion, which precedes the periodicity, is an effective radial current. It is short, spatially finite and unique for each new ion, but multiplied by the rate of ionization and it can produce a substantial total radial current. The associated torque induces rotation which modify the transport processes. We derive the magnitude of the radial current induced by ionization by three methods: the analysis of a simple physical picture, a numerical model and the neoclassical drift-kinetic treatment. The results of the three approaches are in agreement and show that the current can indeed be substantial. Many well known experimental observations can be reconsi...

  5. Concept design on RH maintenance of CFETR Tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Wu, Songtao; Wan, Yuanxi; Li, Jiangang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Ye, Minyou [University of Science and Technology of China, Hefei (China); Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2014-10-15

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.

  6. Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

    Energy Technology Data Exchange (ETDEWEB)

    Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)

    2016-10-15

    Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

  7. Emissive limiter bias experiment for improved confinement of tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Choe, W.; Ono, M.; Darrow, D.S. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Pribyl, P.A.; Liberati, J.R.; Taylor, R.J. (California Univ., Los Angeles, CA (United States). Tokamak Fusion Lab.)

    1992-01-01

    Experiments have been performed in Ohmic discharges of the UCLA CCT tokamak with a LaB[sub 6] biased limiter, capable of emitting energetic electrons as a technique to improve confinement in tokamaks. To study the effects of emitted electrons, the limiter position, bias voltage, and plasma position were varied. The results have shown that the plasma positioning with respect to the emissive limiter plays an important role in obtaining H-mode plasmas. The emissive cathode must be located close to the last closed flux surface in order to charge up the plasma. As the cathode is moved closer to the wall, the positioning of the plasma becomes more critical since the plasma can easily detach from the cathode and reattach to the wall, resulting in the termination of H-mode. The emissive capability appears to be important for operating at lower bias voltage and reducing impurity levels in the plasma. With a heated cathode, transition to H-mode was observed for V[sub bias] [le] 50 V and I[sub inj] [ge] 30 A. At a lower cathode heater current, a higher bias voltage is required for the transition. Moreover, with a lower cathode heater current, the time delay for inducing H-mode becomes longer, which can be attributed to the required time for the self-heating of the cathode to reach the emissive temperature. From this result, we conclude that the capacity for emission can significantly improve the performance of limiter biasing for inducing H-mode transition. With L-mode plasmas, the injection current flowing out of the cathode was generally higher than 100 A.

  8. Emissive limiter bias experiment for improved confinement of tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Choe, W.; Ono, M.; Darrow, D.S. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Pribyl, P.A.; Liberati, J.R.; Taylor, R.J. [California Univ., Los Angeles, CA (United States). Tokamak Fusion Lab.

    1992-10-01

    Experiments have been performed in Ohmic discharges of the UCLA CCT tokamak with a LaB{sub 6} biased limiter, capable of emitting energetic electrons as a technique to improve confinement in tokamaks. To study the effects of emitted electrons, the limiter position, bias voltage, and plasma position were varied. The results have shown that the plasma positioning with respect to the emissive limiter plays an important role in obtaining H-mode plasmas. The emissive cathode must be located close to the last closed flux surface in order to charge up the plasma. As the cathode is moved closer to the wall, the positioning of the plasma becomes more critical since the plasma can easily detach from the cathode and reattach to the wall, resulting in the termination of H-mode. The emissive capability appears to be important for operating at lower bias voltage and reducing impurity levels in the plasma. With a heated cathode, transition to H-mode was observed for V{sub bias} {le} 50 V and I{sub inj} {ge} 30 A. At a lower cathode heater current, a higher bias voltage is required for the transition. Moreover, with a lower cathode heater current, the time delay for inducing H-mode becomes longer, which can be attributed to the required time for the self-heating of the cathode to reach the emissive temperature. From this result, we conclude that the capacity for emission can significantly improve the performance of limiter biasing for inducing H-mode transition. With L-mode plasmas, the injection current flowing out of the cathode was generally higher than 100 A.

  9. Transport bifurcation induced by sheared toroidal flow in tokamak plasmasa)

    Science.gov (United States)

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

    2011-10-01

    First-principles numerical simulations are used to describe a transport bifurcation in a differentially rotating tokamak plasma. Such a bifurcation is more probable in a region of zero magnetic shear than one of finite magnetic shear, because in the former case the component of the sheared toroidal flow that is perpendicular to the magnetic field has the strongest suppressing effect on the turbulence. In the zero-magnetic-shear regime, there are no growing linear eigenmodes at any finite value of flow shear. However, subcritical turbulence can be sustained, owing to the existence of modes, driven by the ion temperature gradient and the parallel velocity gradient, which grow transiently. Nonetheless, in a parameter space containing a wide range of temperature gradients and velocity shears, there is a sizeable window where all turbulence is suppressed. Combined with the relatively low transport of momentum by collisional (neoclassical) mechanisms, this produces the conditions for a bifurcation from low to high temperature and velocity gradients. A parametric model is constructed which accurately describes the combined effect of the temperature gradient and the flow gradient over a wide range of their values. Using this parametric model, it is shown that in the reduced-transport state, heat is transported almost neoclassically, while momentum transport is dominated by subcritical parallel-velocity-gradient-driven turbulence. It is further shown that for any given input of torque, there is an optimum input of heat which maximises the temperature gradient. The parametric model describes both the behaviour of the subcritical turbulence (which cannot be modelled by the quasi-linear methods used in current transport codes) and the complicated effect of the flow shear on the transport stiffness. It may prove useful for transport modelling of tokamaks with sheared flows.

  10. Recent results from the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, P.I.

    1998-02-01

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

  11. Ideal MHD stability of very high beta tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Chance, M.S.; Jardin, S.C.; Kessel, C.; Manickam, J.; Monticello, D. (Princeton Univ., NJ (USA). Plasma Physics Lab.); Peng, Y.K.M.; Holmes, J.A.; Strickler, D.J.; Whitson, J.C. (Oak Ridge National Lab., TN (USA)); Glasser, A.H. (Los Alamos National Lab., NM (USA)); Sykes, A. (UKAEA Culham Lab., Abingdon (UK)); Ramos, J.J. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Plasma Fusion Center)

    1990-12-01

    Achieving very high {beta} and high {beta}{sub p} simultaneously in tokamaks generally implies that the second stability region against ballooning modes must be accessed. We describe several approaches for doing this, which are characterized by the choice of constraints imposed on the equilibrium profiles and the cross-sectional shape of the plasma. The combination of high toroidal beta, restricting the current density to vanish at the edge of the plasma and maintaining a monotonic q profile, proves to be the most stringent. Consideration of equilibria with high {epsilon}{beta}{sub p} but low {beta} facilitates accessibility with peaked pressure profiles and high values of q{sub 0}. Allowing the pressure gradient and, hence, the current density to be finite at the plasma edge allows all surfaces to lie within the second stability regime. For free boundary plasmas with divertors, the divertor stabilized edge region remains in the first stability regime while the plasma core reaches into the second regime. Careful tailoring of the profiles must be used to traverse the unstable barrier commonly seen near the edge of these plasmas. The CAMINO code allows us to compute s-{alpha} curves for general tokamak geometry. These diagrams enable us to construct equilibria whose profiles are only constrained, at worst, to be marginally stable everywhere, but do not necessarily satisfy the constraints on the current or {beta}. There are theoretical indications that under certain conditions the external kinks possess a second region of stability at high q{sub 0} that is analogous to that of the ballooning modes. It is found that extremely accurate numerical means must be developed and applied to confidently establish the validity of these results. 14 refs., 5 figs., 1 tab.

  12. EDICAM fast video diagnostic installation on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Szappanos, A., E-mail: szappanos@rmki.kfki.h [KFKI-RMKI, EURATOM Association, PO Box 49, Budapest-114, H-1521 Budapest (Hungary); Berta, M. [Szechenyi Istvan University, EURATOM Association, Egyetem ter 1, 9026 Gyor (Hungary); Hron, M.; Panek, R.; Stoeckel, J. [Institute of Plasma Physics AS CR, Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Tulipan, S.; Veres, G. [KFKI-RMKI, EURATOM Association, PO Box 49, Budapest-114, H-1521 Budapest (Hungary); Weinzettl, V. [Institute of Plasma Physics AS CR, Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Zoletnik, S. [KFKI-RMKI, EURATOM Association, PO Box 49, Budapest-114, H-1521 Budapest (Hungary)

    2010-07-15

    A new camera system 'event detection intelligent camera' (EDICAM) is being developed by the Hungarian Association and has been installed on the COMPASS tokamak in the Institute of Plasma Physics AS CR in Prague, during February 2009. The standalone system contains a data acquisition PC and a prototype sensor module of EDICAM. Appropriate optical system have been designed and adjusted for the local requirements, and a mechanical holder keeps the camera out of the magnetic field. The fast camera contains a monochrome CMOS sensor with advanced control features and spectral sensitivity in the visible range. A special web based control interface has been implemented using Java spring framework to provide the control features in a graphical user environment. Java native interface (JNI) is used to reach the driver functions and to collect the data stored by direct memory access (DMA). Using a built in real-time streaming server one can see the live video from the camera through any web browser in the intranet. The live video is distributed in a Motion Jpeg format using real-time streaming protocol (RTSP) and a Java applet have been written to show the movie on the client side. The control system contains basic image processing features and the 3D wireframe of the tokamak can be projected to the selected frames. A MatLab interface is also presented with advanced post processing and analysis features to make the raw data available for high level computing programs. In this contribution all the concepts of EDICAM control center and the functions of the distinct software modules are described.

  13. Bulk ion heating with ICRF waves in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Mantsinen, M. J., E-mail: mervi.mantsinen@bsc.es [Catalan Institution for Research and Advanced Studies, Barcelona (Spain); Barcelona Supercomputing Center, Barcelona (Spain); Bilato, R.; Bobkov, V. V.; Kappatou, A.; McDermott, R. M.; Odstrčil, T.; Tardini, G.; Bernert, M.; Dux, R.; Maraschek, M.; Noterdaeme, J.-M.; Ryter, F.; Stober, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Nocente, M. [Dipartimento di Fisica “G. Occhialini”, Università degli Studi di Milano-Bicocca, Milano (Italy); Istituto di Fisica del Plasma “P. Caldirola”, CNR, Milano (Italy); Hellsten, T. [Dept. of Fusion Plasma Physics, EES, KTH, Stockholm (Sweden); Mantica, P.; Tardocchi, M. [Istituto di Fisica del Plasma “P. Caldirola”, CNR, Milano (Italy); Nielsen, S. K.; Rasmussen, J.; Stejner, M. [Technical University of Denmark, Department of Physics, Lyngby (Denmark); and others

    2015-12-10

    Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER and DEMO operation. This is of particular importance for the bulk ion heating capabilities of ICRF waves. Efficient bulk ion heating with the standard ITER ICRF scheme, i.e. the second harmonic heating of tritium with or without {sup 3}He minority, was demonstrated in experiments carried out in deuterium-tritium plasmas on JET and TFTR and is confirmed by ICRF modelling. This paper focuses on recent experiments with {sup 3}He minority heating for bulk ion heating on the ASDEX Upgrade (AUG) tokamak with ITER-relevant all-tungsten PFCs. An increase of 80% in the central ion temperature T{sub i} from 3 to 5.5 keV was achieved when 3 MW of ICRF power tuned to the central {sup 3}He ion cyclotron resonance was added to 4.5 MW of deuterium NBI. The radial gradient of the T{sub i} profile reached locally values up to about 50 keV/m and the normalized logarithmic ion temperature gradients R/LT{sub i} of about 20, which are unusually large for AUG plasmas. The large changes in the T{sub i} profiles were accompanied by significant changes in measured plasma toroidal rotation, plasma impurity profiles and MHD activity, which indicate concomitant changes in plasma properties with the application of ICRF waves. When the {sup 3}He concentration was increased above the optimum range for bulk ion heating, a weaker peaking of the ion temperature profile was observed, in line with theoretical expectations.

  14. Development and validation of a tokamak skin effect transformer model

    Science.gov (United States)

    Romero, J. A.; Moret, J.-M.; Coda, S.; Felici, F.; Garrido, I.

    2012-02-01

    A lumped parameter, state space model for a tokamak transformer including the slow flux penetration in the plasma (skin effect transformer model) is presented. The model does not require detailed or explicit information about plasma profiles or geometry. Instead, this information is lumped in system variables, parameters and inputs. The model has an exact mathematical structure built from energy and flux conservation theorems, predicting the evolution and non-linear interaction of plasma current and internal inductance as functions of the primary coil currents, plasma resistance, non-inductive current drive and the loop voltage at a specific location inside the plasma (equilibrium loop voltage). Loop voltage profile in the plasma is substituted by a three-point discretization, and ordinary differential equations are used to predict the equilibrium loop voltage as a function of the boundary and resistive loop voltages. This provides a model for equilibrium loop voltage evolution, which is reminiscent of the skin effect. The order and parameters of this differential equation are determined empirically using system identification techniques. Fast plasma current modulation experiments with random binary signals have been conducted in the TCV tokamak to generate the required data for the analysis. Plasma current was modulated under ohmic conditions between 200 and 300 kA with 30 ms rise time, several times faster than its time constant L/R ≈ 200 ms. A second-order linear differential equation for equilibrium loop voltage is sufficient to describe the plasma current and internal inductance modulation with 70% and 38% fit parameters, respectively. The model explains the most salient features of the plasma current transients, such as the inverse correlation between plasma current ramp rates and internal inductance changes, without requiring detailed or explicit information about resistivity profiles. This proves that a lumped parameter modelling approach can be used to

  15. Neutral particle dynamics in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Niemczewski, A.P.

    1995-08-01

    This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism. 146 refs., 82 figs., 14 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-15

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

  17. Study of heat flux deposition in the Tore Supra Tokamak; Etude des depots de chaleur dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Carpentier, S.

    2009-02-15

    Accurate measurements of heat loads on internal tokamak components is essential for protection of the device during steady state operation. The optimisation of experimental scenarios also requires an in depth understanding of the physical mechanisms governing the heat flux deposition on the walls. The objective of this study is a detailed characterisation of the heat flux to plasma facing components (PFC) of the Tore Supra tokamak. The power deposited onto Tore Supra PFCs is calculated using an inverse method, which is applied to both the temperature maps measured by infrared thermography and to the enthalpy signals from calorimetry. The derived experimental heat flux maps calculated on the toroidal pumped limiter (TPL) are then compared with theoretical heat flux density distributions from a standard SOL-model. They are two experimental observations that are not consistent with the model: significant heat flux outside the theoretical wetted area, and heat load peaking close to the tangency point between the TPL and the last closed field surface (LCFS). An experimental analysis for several discharges with variable security factors q is made. In the area consistent with the theoretical predictions, this parametric study shows a clear dependence between the heat flux length lambda{sub q} (estimated in the SOL (scrape-off layer) from the IR measurements) and the magnetic configuration. We observe that the spreading of heat fluxes on the component is compensated by a reduction of the power decay length lambda{sub q} in the SOL when q decreases. On the other hand, in the area where the derived experimental heat loads are not consistent with the theoretical predictions, we observe that the spreading of heat fluxes outside the theoretical boundary increases when q decreases, and is thus not counterbalanced. (author)

  18. Preliminary project of s Thomson scattering system for the ETE tokamak; Projeto preliminar de um sistema de espalhamento Thomson para o Tokamak ETE

    Energy Technology Data Exchange (ETDEWEB)

    Berni, Luiz Angelo

    1997-12-31

    This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light 4 refs., 26 figs.

  19. Electron density and temperature determination in a Tokamak plasma using light scattering; Determinacion de la densidad y temperatura electronicas en un Tokamak mediante difusion luminosa

    Energy Technology Data Exchange (ETDEWEB)

    Perez-Navarro Gomerz, A.; Zurro Hernandez, B.

    1976-07-01

    A theoretical foundation review for light scattering by plasmas is presented. Furthermore, we have included a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, in a Tokamak plasma using spectral analysis of the scattered radiation. (Author) 13 refs.

  20. Upgraded ECE radiometer on the Tore Supra Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Segui, J.L.; Molina, D.; Goniche, M.; Maget, P.; Udintsev, V.S. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Antar, G.Y. [Center for Energy Research, UCSD, La Jolla CA (United States); Kraemer-Flecken, A. [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Plasmaphysik

    2004-07-01

    An upgraded 32-channel heterodyne radiometer, 1 GHz spaced, is used on the Tore-Supra tokamak to measure the electron cyclotron emission (ECE) in the frequency range 78-110 GHz for the ordinary mode (O1) and 94-126.5 GHz for the extraordinary mode (X2). From now radial resolution is essentially limited by ECE relativistic effects related to electron temperature and density, not by the channels frequency spacing. For example, this leads to precise electron temperature mapping during magneto hydrodynamic activities (MHD). In the equatorial plane, we use a dual polarisation Gaussian optics lens antenna. It has low spreading and a perpendicular line-of-sight that gives ECE measurements very low refraction and Doppler effects. Assuming that the plasma is a black body and there is no overlap between ECE harmonics, one can deduce the electron temperature profile by using the first harmonic ordinary mode (O1) or the second harmonic extraordinary mode (X2). The principle radio frequency emitter (RF) has its frequencies down shifted into intermediary frequencies (IF) that span from 2 to 18 GHz in the single side band mode (SSB). It is amplified by low noise IF amplifiers before forming channels. A separate O/X mode RF front-end allows the use of an IF electronic mode selector. This gives the potentiality of simultaneous O/X mode measurements in the 94-110 GHz. RF and IF filters reject the gyrotron frequency (118 GHz) in order to perform electron temperature measurements during electron cyclotron resonance heated plasmas. A precise absolute spectral calibration is performed outside the tokamak vacuum vessel by using a 600 deg C black body hot source, a double coherent digital signal averaging (trigger, turn and clock) on the waveform generated by a mechanical chopper, and a simulated tokamak window. The use of differential electronics and strong electromagnetic shielding improves also the calibration precision. The fast and slow data acquisition systems are free of aliasing

  1. Density limits investigation and high density operation in EAST tokamak

    Science.gov (United States)

    Zheng, Xingwei; Li, Jiangang; Hu, Jiansheng; Liu, Haiqing; Jie, Yinxian; Wang, Shouxin; Li, Jiahong; Duan, Yanming; Li, Miaohui; Li, Yongchun; Zhang, Ling; Ye, Yang; Yang, Qingquan; Zhang, Tao; Cheng, Yingjie; Xu, Jichan; Wang, Liang; Xu, Liqing; Zhao, Hailin; Wang, Fudi; Lin, Shiyao; Wu, Bin; Lyu, Bo; Xu, Guosheng; Gao, Xiang; Shi, Tonghui; He, Kaiyang; Lan, Heng; Chu, Nan; Cao, Bin; Sun, Zhen; Zuo, Guizhong; Ren, Jun; Zhuang, Huidong; Li, Changzheng; Yuan, Xiaolin; Yu, Yaowei; Wang, Houyin; Chen, Yue; Wu, Jinhua; EAST Team

    2016-05-01

    Increasing the density in a tokamak is limited by the so-called density limit, which is generally performed as an appearance of disruption causing loss of plasma confinement, or a degradation of high confinement mode which could further lead to a H  →  L transition. The L-mode and H-mode density limit has been investigated in EAST tokamak. Experimental results suggest that density limits could be triggered by either edge cooling or excessive central radiation. The L-mode density limit disruption is generally triggered by edge cooling, which leads to the current profile shrinkage and then destabilizes a 2/1 tearing mode, ultimately resulting in a disruption. The L-mode density limit scaling agrees well with the Greenwald limit in EAST. The observed H-mode density limit in EAST is an operational-space limit with a value of 0.8∼ 0.9{{n}\\text{GW}} . High density H-mode heated by neutral beam injection (NBI) and lower hybrid current drive (LHCD) are analyzed, respectively. The constancy of the edge density gradients in H-mode indicates a critical limit caused perhaps by e.g. ballooning induced transport. The maximum density is accessed at the H  →  L transition which is generally caused by the excessive core radiation due to high Z impurities (Fe, Cu). Operating at a high density (>2.8× {{10}19} {{\\text{m}}-3} ) is favorable for suppressing the beam shine through NBI. High density H-mode up to 5.3× {{10}19}{{\\text{m}}-3}~≤ft(∼ 0.8{{n}\\text{GW}}\\right) could be sustained by 2 MW 4.6 GHz LHCD alone, and its current drive efficiency is studied. Statistics show that good control of impurities and recycling facilitate high density operation. With careful control of these factors, high density up to 0.93{{n}\\text{GW}} stable H-mode operation was carried out heated by 1.7 MW LHCD and 1.9 MW ion cyclotron resonance heating with supersonic molecular beam injection fueling.

  2. Magnetic spires for the detection of the position of the plasma column in a Tokamak (linear approximation); Espiras magneticas para la deteccion de la posicion de la columna de plasma en un Tokamak (aproximacion lineal)

    Energy Technology Data Exchange (ETDEWEB)

    Colunga S, S

    1990-07-15

    In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)

  3. Optimization of magnetic field system for glass spherical tokamak GLAST-III

    Science.gov (United States)

    Ahmad, Zahoor; Ahmad, S.; Naveed, M. A.; Deeba, F.; Aqib Javeed, M.; Batool, S.; Hussain, S.; Vorobyov, G. M.

    2017-04-01

    GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA-1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils.

  4. A new trial of tokamak in-vessel inspection manipulator

    Energy Technology Data Exchange (ETDEWEB)

    Du, Liang; Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Zhang, Weijun; Li, Fashe

    2015-10-15

    In this paper, we discuss the design and partial implementation of an in-vessel inspection manipulator in detail, which is considered to serve for China's Experimental Advanced Superconducting Tokamak (EAST). Besides the ordinary kinematic/dynamic constraints and specifications for a multiple degrees of freedom (DOF) manipulator suitable for EAST in-vessel inspection, there is extra necessity in design for the extreme in-vessel environment, e.g., high temperature and high vacuum. Based on our recent developed active cooling system, a specific proposal is explored, which employs ordinary commercial mechanical/electrical components only, as if the manipulator works in normal temperature environment. This paper also emphasizes some challenging technical issues toward an implementation, such as an optimization of thermal gradient/cooling path in the manipulator, a trade-off between large reachable space and large rotation angle of each joint, a special designed revolute joint structure for cooling tube arrangement and so on. We use an EtherCAT based real time control platform connecting drivers and sensors, which achieves a robust closed-loop system and a clean cable aspect simultaneously. In the later part of the paper, basic mechanical tests and inspection process are described. Evaluation on recent progress and future work toward a whole-scale test is stated and expected.

  5. Discharge cleaning and wall conditioning in a Novillo Tokamak

    CERN Document Server

    Valencia, R; Camps, E; Contreras, G; Muhl, S

    2002-01-01

    Our Novillo Tokamak is a small toroidal device magnetically confined defined by the main design parameters: R sub o =0.23 m, a sub v =0.08 m, a sub p =0.06 m, B sub T =0.05-0.47 T, I sub p =1-12 kA, n sub e =1-2x10 sup 1 sup 3 cm sup - sup 3 , T sub e =150 eV, T sub i =50 eV. For the initial discharge chamber cleaning we have often used vacuum baking up to 100 deg. C and then conditioning using Taylor discharge cleaning (TDC) in H sub 2 and He. In this work we report that vacuum baking is effective for obtaining a final total pressure of the order of 1.6x10 sup - sup 7 Torr. We have found that a single parameter, the performance parameter (PP), can be used to optimize the TDC method. This parameter represents the quantity of electron and ion energy incident on the chamber wall during the Taylor discharge, it is equal to (I sub p tau), where I sub p is the peak-to-peak plasma current and tau is the plasma current duration. In graphs of PP versus the gas pressure for different oscillator powers, the maximum val...

  6. Upgraded data service system for HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    QU Lian-Zheng; LUO Jia-Rong; WEI Pei-Jie; LI Gui-Ming; CHENG Ting; QI Na

    2005-01-01

    A data service system plays an indispensable role in HT-7 Tokamak experiment. Since the former system doesn't provide the function of timely data procession and analysis, and all client software are based on Windows, it can't fulfill virtual fusion laboratory for remote researchers. Therefore, a new system which is simplified by three kinds of data servers and one data analysis and visualization software tool has been developed. The data servers include a data acquisition server based on file system, an MDSplus server used as the central repository for analysis data, and a web server. Users who prefer the convenience of application that can be run in a Web Browser can easily access the experiment data without knowing X-Windows. In order to adjust instruments to control experiment the operators need to plot data duly as soon as they are gathered. To satisfy their requirement, an upgraded data analysis and visualization software GT-7 is developed. It not only makes 2D data visualization more efficient, but also it can be capable of processing, analyzing and displaying interactive 2D and 3D graph of raw, analyzed data by the format of ASCII, LZO and MDSplus.

  7. Modernized active spectroscopic diagnostics (CXRS) of the T-10 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Krupin, V. A., E-mail: Krupin-VA@nrcki.ru; Klyuchnikov, L. A., E-mail: Lklyuchnikov@list.ru; Korobov, K. V., E-mail: Korobov-KV@nrcki.ru; Nemets, A. R., E-mail: Nemets-AR@rncki.ru; Nurgaliev, M. R.; Gorbunov, A. V. [National Research Center Kurchatov Institute (Russian Federation); Naumenko, N. N. [National Academy of Sciences of Belarus, Stepanov Institute of Physics (Belarus); Troynov, V. I.; Tugarinov, S. N.; Fomin, F. V. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    This work presents the results of modernization of the CXRS (charge exchange recombination spectroscopy) diagnostics [1] at the T-10 tokamak. The relevance of this work is due to the importance of measurements of the ion temperature and nuclei density of the working gas and impurities for analysis of transport processes in the plasma ion component. Measurements of radial profiles of the ion temperature are extremely important for investigating the geodesic acoustic mode behavior which is conducted at the T-10 [2]. The modernized scheme of CXRS measurements, as well as the design and operational features of the spectrometer created for the new diagnostics, is described. Principles of data recording and further processing are considered in detail; attention is given to the problem of calibration of the whole complex of equipment. The performed changes in diagnostics allow the measurements to be taken simultaneously in three spectral intervals: in the region of the beam line H{sub α}, the CXRS line of carbon ion C{sup 5+}, and the CXRS line of one of the hydrogen-like ions: He{sup 1+}, Li{sup 2+}, N{sup 6+}, O{sup 7+} or Ne{sup 9+}. This makes it possible to measure the density profiles of two plasma impurities simultaneously, as well as the ion temperature from CXRS lines of different elements. The modernized diagnostics significantly broadens the possibilities of studying the physics of transport processes and quasi-coherent modes of plasma oscillations at the T-10.

  8. ADVANCES IN COMPREHENSIVE GYROKINETIC SIMULATIONS OF TRANSPORT IN TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    WALTZ RE; CANDY J; HINTON FL; ESTRADA-MILA C; KINSEY JE

    2004-10-01

    A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate core turbulent transport in actual experimental profiles and enable direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite {beta}, equilibrium ExB shear stabilization, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius ({rho}{sub *}) so as to treat the profile shear stabilization and nonlocal effects which can break gyroBohm scaling. The code operates in either a cyclic flux-tube limit (which allows only gyroBohm scaling) or a globally with physical profile variation. Rohm scaling of DIII-D L-mode has been simulated with power flows matching experiment within error bars on the ion temperature gradient. Mechanisms for broken gyroBohm scaling, neoclassical ion flows embedded in turbulence, turbulent dynamos and profile corrugations, plasma pinches and impurity flow, and simulations at fixed flow rather than fixed gradient are illustrated and discussed.

  9. Kinetic modelling of runaway electron avalanches in tokamak plasmas

    CERN Document Server

    Nilsson, E; Peysson, Y; Granetz, R S; Saint-Laurent, F; Vlainic, M

    2015-01-01

    Runaway electrons (REs) can be generated in tokamak plasmas if the accelerating force from the toroidal electric field exceeds the collisional drag force due to Coulomb collisions with the background plasma. In ITER, disruptions are expected to generate REs mainly through knock-on collisions, where enough momentum can be transferred from existing runaways to slow electrons to transport the latter beyond a critical momentum, setting off an avalanche of REs. Since knock-on runaways are usually scattered off with a significant perpendicular component of the momentum with respect to the local magnetic field direction, these particles are highly magnetized. Consequently, the momentum dynamics require a full 3-D kinetic description, since these electrons are highly sensitive to the magnetic non-uniformity of a toroidal configuration. A bounce-averaged knock-on source term is derived. The generation of REs from the combined effect of Dreicer mechanism and knock-on collision process is studied with the code LUKE, a s...

  10. Sideways force due to coupled kink modes in tokamaks

    Science.gov (United States)

    Mironov, D. V.; Pustovitov, V. D.

    2017-09-01

    The paper is devoted to calculation of the sideways force on the vacuum vessel wall in tokamaks at the early stage of disruptions before possible appearance of the halo currents. The theory developed so far predicts quite a large force due to a single-helicity kink mode. This concept is revisited here. It is demonstrated that the existing single-mode models are incompatible with the natural requirement that the sideways force on the plasma must be practically zero. The latter can be satisfied by incorporating a secondary coupled mode, which is the main new element here compared to the earlier analytical approaches. With this difference, the derivations are performed in the standard cylindrical model that is widely used in the resistive wall mode studies. The resistive wall effects are accounted for as determining the wall reaction on the plasma-driven perturbations. The derived expressions explicitly reveal the sideways force dependence on γτw with a maximum at γτw=O (1 ) , where γ is the kink growth rate and τw is the resistive wall time. It is proved that the amplitude of the sideways force produced by the kink modes must be much smaller than expected from the existing scalings.

  11. Mechanisms of plasma disruption and runaway electron losses in tokamaks

    CERN Document Server

    Abdullaev, S S; Wongrach, K; Tokar, M; Koslowski, H R; Willi, O; Zeng, L

    2015-01-01

    Based on the analysis of data from the numerous dedicated experiments on plasma disruptions in the TEXTOR tokamak mechanisms of the formation of runaway electron beams and their losses are proposed. The plasma disruption is caused by strong stochastic magnetic field formed due to nonlinearly excited low-mode number MHD modes. It is hypothesized that the runaway electron beam is formed in the central plasma region confined inside the intact magnetic surface located between $q=1$ and the closest low--order rational [$q=4/3$ or $q=3/2$] magnetic surfaces. The thermal quench time caused by the fast electron transport in a stochastic magnetic field is calculated using the collisional transport model. The current decay stage is due to the ambipolar particle transport in a stochastic magnetic field. The runaway electron beam in the confined plasma region is formed due to their acceleration the inductive toroidal electric field. The runaway electron beam current is modeled as a sum of toroidally symmetric part and a ...

  12. Mechanisms of Extending Operation Regionin the HL-1M Tokamak

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Stable operating region in the HL-1M tokamak has been extended by means ofwall conditioning, core fuelling and current control techniques. The mechanisms of the extensionare analyzed in this paper. Lithiumization diminishes the impurities and hydrogen recycling tothe lowest level. After lithiumization a high density up to 7×1019 m-3 was obtained easily bystrong gas puffing with ordinary ohmic discharge alone. More attractively we found that metalLi-coating exhibited the effects of wall stabilization. The low qa limit with higher density wasextended by a factor of 1.5~2 in comparison with that for boronization, and 1.2 for siliconization.Siliconization not only extended stable operating region significantly by itself, but also provideda good target plasma for other experiments of raising density limit. Core fuelling schemes arefavourable especially for siliconized wall with a higher level of medium-Z impurity (Z=14).After siliconization the maximum density near to 1020 m-3 was achieved by a combination ofsupersonic molecule beam injection and multipellet injection. The new defined slope of Hugilllimit illustrating more clearly the situation under low qa and high ne discharges was created toindicate the new region extended by combining Ip ramp-up with core fuelling. The slope with alarge Murakami coefficient increased by a factor of 50~60 %.

  13. Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

    Energy Technology Data Exchange (ETDEWEB)

    Xu, X Q

    2007-11-09

    We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D ({psi}, {theta}, {epsilon}, {mu}) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices.

  14. A novel compact Tokamak Hard X-ray diagnostic detector

    Institute of Scientific and Technical Information of China (English)

    曹靖; 蒋春雨; 赵艳凤; 杨青巍; 阴泽杰

    2015-01-01

    A compact X-ray detector based on the lutetium yttrium oxyorthosilicate scintillator (LYSO) and silicon photomultiplier (SiPM) has been designed and fabricated for the hard X-ray diagnosis on the HL 2A and HL 2M Tokamak devices. The LYSO scintillator and SiPM in small dimensions were combined in a heat shrink tube package, making the detector compact and integrative. The Monte Carlo particle transport simulation tool, Geant4, was utilized for the design of the detector for the hard X-ray from 10 keV to 200 keV and the best structure scheme was presented. Finally, the detector was used to measure the photon spectrum of a 137Cs gamma source with a pre-amplifier and a multichannel amplitude analyzer. The measured spectrum is consistent with the theoretic spectrum, it has shown that the energy resolution of the detector is less than 14.8%at an energy of 662 keV.

  15. What sets the minimum tokamak scrape-off layer width?

    Science.gov (United States)

    Joseph, Ilon

    2016-10-01

    The heat flux width of the tokamak scrape-off layer is on the order of the poloidal ion gyroradius, but the ``heuristic drift'' physics model is still not completely understood. In the absence of anomalous transport, neoclassical transport sets the minimum width. For plateau collisionality, the ion temperature width is set by qρi , while the electron temperature width scales as the geometric mean q(ρeρi) 1 / 2 and is close to qρi in magnitude. The width is enhanced because electrons are confined by the sheath potential and have a much longer time to radially diffuse before escaping to the wall. In the Pfirsch-Schluter regime, collisional diffusion increases the width by the factor (qR / λ) 1 / 2 where qR is the connection length and λ is the mean free path. This qualitatively agrees with the observed transition in the scaling law for detached plasmas. The radial width of the SOL electric field is determined by Spitzer parallel and ``neoclassical'' radial electric conductivity and has a similar scaling to that for thermal transport. Prepared under US DOE contract DE-AC52-07NA27344.

  16. Vacuum vessel system design for the compact ignition tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reddan, W. (Ebasco Services Inc., Princeton, NJ (USA))

    1990-05-01

    The compact ignition tokamak (CIT) is envisioned to be the test bed for the study of self- sustained, or ignited, fusion plasmas. The design basis for CIT is a 11-T toroidal field, 12-MA plasma current and peak fusion power of 500 MW. A major portion of this project is the vacuum vessel system, which includes the vacuum chamber, the divertor, first wall, and the robotics systems necessary to maintain the in-vessel components. The vacuum chamber is 2.1 m major radius torus with a D-shaped cross section. For hydrogenic species the base pressure is 10{sup {minus}7} Torr, with a total pumping speed of 5000 l/s. It is designed to withstand the forces resulting from plasma disruptions and be bakeable to approximately 350 {degree}C. A swept divertor and fixed limiters are provided. Both are carbon based structures designed to accommodate heat fluxes as large as 40 MW/m{sup 2} during the 5 s pulse. Articulated booms and manipulators will be deployed for in-vessel maintenance tasks, such as first wall removal/replacement and leak checking. This paper summarizes the engineering considerations and design status. In addition, the unique organization of the project's national design team, led by the Princeton Plasma Physics Laboratory, and the integration into this organization of the industrial consortium responsible for the design and fabrication of the vacuum vessel system is described.

  17. Oak Ridge Tokamak experimental power reactor study scoping report

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, M.

    1977-03-01

    This report presents the scoping studies performed as the initial part of the program to produce a conceptual design for a Tokamak Experimental Power Reactor (EPR). The EPR as considered in this study is to employ all systems necessary for significant electric power production at continuous high duty cycle operation; it is presently scheduled to be the final technological step before a Demonstration Reactor Plant (Demo). The scoping study tasks begin with an exploration and identification of principal problem areas and then concentrate on consideration and evaluation of alternate design choices for each of the following major systems: Plasma Engineering and Physics, Nuclear, Electromagnetics, Neutral Beam Injection, and Tritium Handling. In addition, consideration has been given to the integration of these systems and requirements arising out of their incorporation into an EPR. One intent of this study is to document the paths explored in search of the appropriate EPR characteristics. To satisfy this intent, the explorations are presented in chart form outlining possible options in key areas with extensive supporting footnotes. An important result of the scoping study has been the development and definition of an EPR reference design to serve as (1) a common focus for the continuing design study and (2) a guide for associated development programs. In addition, the study has identified research and development requirements essential to facilitate the successful conceptual design, construction, and operation of an EPR.

  18. Study on wall recycling behaviour in CPD spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyay, R. [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)], E-mail: raju@triam.kyushu-u.ac.jp; Zushi, H. [Research Institute of Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Hirooka, Y. [National Institute for Fusion Science, Toki 509-5292 (Japan); Sakamoto, M. [Research Institute of Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Yoshinaga, T. [National Institute for Fusion Science, Toki 509-5292 (Japan); Okamoto, K. [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Kawasaki, S.; Hanada, K.; Sato, K.N.; Nakamura, K.; Idei, H. [Research Institute of Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Ryoukai, T. [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Nakashima, H.; Higashijima, A. [Research Institute of Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2008-12-15

    Experiments to study wall recycling behaviour have been performed in the small spherical tokamak compact plasma-wall interaction experimental device (CPD) from the viewpoint of global as well as local plasma wall interaction condition. Electron cyclotron resonance (ECR) plasma of typically {approx}50 to 400 ms duration is produced using {approx}40 to 80 kW RF power. In order to study the global wall recycling behaviour, pressure measurements are carried out just before and after the ECR plasma in the absence of any external pumping. The recycling behaviour is found to change from release to pumping beyond a certain level of pressure value which is again found to be a function of shot history. The real-time local wall behaviour is studied in similar RF plasma using a rotating tungsten limiter, actively coated with lithium. Measurement of H{sub {alpha}} light intensity in front of the rotating surface has indicated a clear reduction ({approx}10%) in the steady-state hydrogen recycling with continuous Li gettering of several minutes.

  19. Advanced fusion technologies developed for JT-60 superconducting tokamak

    Science.gov (United States)

    Sakasai, A.; Ishida, S.; Matsukawa, M.; Akino, N.; Ando, T.; Arai, T.; Ezato, K.; Hamada, K.; Ichige, H.; Isono, T.; Kaminaga, A.; Kato, T.; Kawano, K.; Kikuchi, M.; Kizu, K.; Koizumi, N.; Kudo, Y.; Kurita, G.; Masaki, K.; Matsui, K.; Miura, Y. M.; Miya, N.; Miyo, Y.; Morioka, A.; Nakajima, H.; Nunoya, Y.; Oikawa, A.; Okuno, K.; Sakurai, S.; Sasajima, T.; Satoh, K.; Shimizu, K.; Takeji, S.; Takenaga, K.; Tamai, H.; Taniguchi, M.; Tobita, K.; Tsuchiya, K.; Urata, K.; Yagyu, J.

    2004-02-01

    Modification of JT-60 as a full superconducting tokamak (JT-60SC) is planned. The objectives of the JT-60SC programme are to establish scientific and technological bases for steady-state operation of high performance plasmas and utilization of reduced-activation materials in an economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to the DEMO reactor have been developed for the superconducting magnet technology and plasma facing components of the JT-60SC design. To achieve a high current density in a superconducting strand, Nb3Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFCs) of JT-60SC. The R&D to demonstrate the applicability of the Nb3Al conductor to TFCs by a react-and-wind technique has been carried out using a full-size Nb3Al conductor. A full-size NbTi conductor with low ac loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the carbon fibre composite target was successfully demonstrated on an electron beam irradiation stand.

  20. Tokamak experimental power reactor conceptual design. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H/sub 2/O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters.

  1. Effects of electrode biasing in STOR-M Tokamak

    Science.gov (United States)

    Basu, Debjyoti; Nakajima, Masaru; Rohollahi, Akbar; McColl, David; Adegun, Joseph; Xiao, Chijin; Hirose, Akira

    2015-11-01

    STOR-M is an iron-core, limiter based tokamak with major and minor radii of 46cm and 12 cm, respectively. Recently, electrode biasing experiments have been carried to study the improved confinement. For this purpose we have developed a DC power supply which can be gated by a high power SCR. The rectangular SS electrode has a height of 10 cm, a width of 2 cm and a thickness of 0.2 cm. The radial position of the electrode throughout the experiments is kept around 4mm inside the limiter in the plasma edge region. After application of positive bias with voltages between +90 V to +110 V during the plasma discharge current flat top with slightly higher edge-qa (nearly 5 to 6), noticeable increment of average plasma density and soft x-ray intensity along the central chord have been observed. No distinguishable change in H α emission has been measured. These phenomena may be attributed to improved confinement formed at the inner region but not at the edge. In the upcoming experimental campaign, Ion Doppler spectroscopy will be used to measure possible velocity shear inside the inner plasma region. Edge plasma pressure gradient will also be measured using Langmuir probes. Detailed experimental results will be presented.

  2. Modelisation of synchrotron radiation losses in realistic tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Albajar, F.; Johner, J.; Granata, G

    2000-08-01

    Synchrotron radiation losses become significant in the power balance of high-temperature plasmas envisaged for next step tokamaks. Due to the complexity of the exact calculation, these losses are usually roughly estimated with expressions derived from a plasma description using simplifying assumptions on the geometry, radiation absorption, and density and temperature profiles. In the present article, the complete formulation of the transport of synchrotron radiation is performed for realistic conditions of toroidal plasma geometry with elongated cross-section, using an exact method for the calculation of the absorption coefficient, and for arbitrary shapes of density and temperature profiles. The effects of toroidicity and temperature profile on synchrotron radiation losses are analyzed in detail. In particular, when the electron temperature profile is almost flat in the plasma center, as for example in ITB confinement regimes, synchrotron losses are found to be much stronger than in the case where the profile is represented by its best generalized parabolic approximation, though both cases give approximately the same thermal energy contents. Such an effect is not included in present approximate expressions. Finally, we propose a seven-variable fit for the fast calculation of synchrotron radiation losses. This fit is derived from a large database, which has been generated using a code implementing the complete formulation and optimized for massively parallel computing. (author)

  3. Experimental observations of driven and intrinsic rotation in tokamak plasmas

    Science.gov (United States)

    Rice, J. E.

    2016-08-01

    Experimental observations of driven and intrinsic rotation in tokamak plasmas are reviewed. For momentum sources, there is direct drive from neutral beam injection, lower hybrid and ion cyclotron range of frequencies waves (including mode conversion flow drive), as well as indirect \\mathbf{j}× \\mathbf{B} forces from fast ion and electron orbit shifts, and toroidal magnetic field ripple loss. Counteracting rotation drive are sinks, such as from neutral drag and toroidal viscosity. Many of these observations are in agreement with the predictions of neo-classical theory while others are not, and some cases of intrinsic rotation remain puzzling. In contrast to particle and heat fluxes which depend on the relevant diffusivity and convection, there is an additional term in the momentum flux, the residual stress, which can act as the momentum source for intrinsic rotation. This term is independent of the velocity or its gradient, and its divergence constitutes an intrinsic torque. The residual stress, which ultimately responds to the underlying turbulence, depends on the confinement regime and is a complicated function of collisionality, plasma shape, and profiles of density, temperature, pressure and current density. This leads to the rich intrinsic rotation phenomenology. Future areas of study include integration of these many effects, advancement of quantitative explanations for intrinsic rotation and development of strategies for velocity profile control.

  4. EBT: an alternate concept to tokamaks and mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Glowienka, J.C.

    1980-01-01

    The ELMO Bumpy Torus (EBT) is a hybrid magnetic trap formed by a series of toroidally connected simple mirrors. It differs from a tokamak, the present main-line approach, in that plasma stability and heating are obtained in a current-free geometry by the application of steady-state, high power, electron cyclotron resonance heating (ECH) producing a steady-state plasma. The primary motivation for EBT confinement research is the potential for a steady-state, highly accessible reactor with high ..beta... In the present EBT-I/S device, electron confinement has been observed to agree with the predictions of theory. The major emphasis of the experimental program is on the further scaling of plasma parameters in the EBT-I/S machine with ECH frequency (10.6, 18, and 28 GHz), resonant magnetic field (0.3, 0.6, and 1 T), and heating power (30, 60, and 200 kW). In addition, substantial efforts are under way or planned in the areas of ion cyclotron heating, neutral beam heating, plasma-wall interactions, impurity control, synchrotron radiation, and divertors. Recently, EBT has been selected as the first alternative concept to be advanced to the proof-of-principle stage; this entails a major device scale-up to allow a reasonable extrapolation to a DT-burning facility. The status and future plans of the EBT program, in particular the proof-of-principle experiment (EBT-P), are discussed.

  5. Linear Analysis of Drift Ballooning Modes in Tokamak Edge Plasmas

    Science.gov (United States)

    Tangri, Varun; Kritz, Arnold; Rafiq, Tariq; Pankin, Alexei

    2012-10-01

    The H-mode pedestal structure depends on the linear stability of drift ballooning modes (DBMs) in many H-mode pedestal models. Integrated modeling that uses these pedestal models requires fast evaluation of linear stability of DBMs. Linear analysis of DBMs is also needed in the computations of effective diffusivities associated with anomalous transport that is driven by the DBMs in tokamak edge plasmas. In this study several numerical techniques of linear analysis of the DBMs are investigated. Differentiation matrix based spectral methods are used to compute the physical eigenvalues of the DBMs. The model for DBMs used here consists of six differential equations [T. Rafiq et al. Phys. Plasmas, 17, 082511, (2010)]. It is important to differentiate among non-physical (numerical) modes and physical modes. The determination of the number of eigenvalues is solved by a computation of the `nearest' and `ordinal' distances. The Finite Difference, Hermite and Sinc based differentiation matrices are used. It is shown that spectral collocation methods are more accurate than finite difference methods. The technique that has been developed for calculating eigenvalues is quite general and is relevant in the computation of other modes that utilize the ballooning mode formalism.

  6. The Alignment and Assembly for EAST Tokamak Device

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    EAST (HT-7U) is a large fusion experimental device. It is a full superconducting tokamak with 1 MA of plasma current, 1000 s of plasma duration, high elongation and triangularity. It mainly consists of superconducting magnets of poloidal and toroidal field (PF& TF),vacuum vessel (VV), thermal radiation shield (TRS) and cryostat vessel (CV). The significant difficulty for assembly of EAST is tight installation tolerances, which are in the order of several tenth of a millimeter. In particular, the alignment of plasma facing components to the magnetic axis of the device is less than ± 0.5 mm.At present, a reasonable assembly process of EAST has been defined, and based on it, the alignment method for EAST, including the survey control network, the location of the main components in different directions, the magnetic axis determination and the accurate positioning of the plasma facing components inside of the vacuum vessel and so on, has been defined by using the sophisticated optical metrology system (SOMS).This paper describes the assembly procedure of EAST and the installation tolerances associated with the main components. Meanwhile, how to establish the assembly survey control network,magnetic axis determination methods, are introduced in detail.

  7. Enhanced Lower Hybrid Current Drive Experiments on HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    Effective Lower Hybrid Current Driving (LHCD) and improved confinement exper-iments in higher plasma parameters (Ip > 200 kA, ne> 2×1013 cm-3, Te ≥ 1 keⅤ) havebeen curried out in optimized LH wave spectrum and plasma parameters in HT-7 supercon-ducting tokamak. The dependence of current driving efficiency on LH power spectrum, plasmadensity ne and toroidal magnetic field BT has been obtained under optimal conditions. A goodCD efficiency was obtained at higher plasma current and higher electron density. The improve-ment of the energy confinement time is accompanied with the increase in line averaged electrondensity, and in ion and electron temperatures. The highest current driving efficiency reachedηCD = IpneR/PRF ≈ 1.05 × 1019 Am-2/W. Wave-plasma coupling was sustained in a good stateand the reflective coefficient was less than 5%. The experiments have also demonstrated the abilityof LH wave in the start-up and ramp-up of the plasma current. The measurement of the temporaldistribution of plasma parameter shows that lower hybrid leads to a broader profile in plasmaparameter. The LH power deposition profile and the plasma current density profile were modeledwith a 2D Fokker-Planck code corresponding to the evolution process of the hard x-ray detectorarray.

  8. Tritium Removal by Laser Heating and Its Application to Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    C.H. Skinner; C.A. Gentile; G. Guttadora; A. Carpe; S. Langish; K.M. Young; M. Nishi; W. Shu

    2001-11-16

    A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm{sup 2}, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed.

  9. Tokamak equilibria with strong toroidal current density reversal

    Science.gov (United States)

    Ludwig, G. O.; Rodrigues, Paulo; Bizarro, João P. S.

    2013-05-01

    The equilibrium of large magnetic islands in the core of a tokamak under conditions of strong toroidal current density reversal is investigated by a new method. The method uses distinct spectral representations to describe each simply connected region as well as the containing shell geometry. This ideal conducting shell may substitute for the plasma edge region or take a virtual character representing the external equilibrium field effect. The internal equilibrium of the islands is solved within the framework of the variational moment method. Equivalent surface current densities are defined on the boundaries of the islands and on the thin containing shell, giving a straightforward formulation to the interaction between regions. The equilibrium of the island-shell system is determined by matching moments of the Dirichlet boundary conditions. Finally, the macroscopic stability against a class of tilting displacements is examined by means of an energy principle. It is found out that the up-down symmetric islands are stable to this particular perturbation and geometry but the asymmetric system presents a bifurcation in the equilibrium.

  10. In situ ``artificial plasma'' calibration of tokamak magnetic sensors

    Science.gov (United States)

    Shiraki, D.; Levesque, J. P.; Bialek, J.; Byrne, P. J.; DeBono, B. A.; Mauel, M. E.; Maurer, D. A.; Navratil, G. A.; Pedersen, T. S.; Rath, N.

    2013-06-01

    A unique in situ calibration technique has been used to spatially calibrate and characterize the extensive new magnetic diagnostic set and close-fitting conducting wall of the High Beta Tokamak-Extended Pulse (HBT-EP) experiment. A new set of 216 Mirnov coils has recently been installed inside the vacuum chamber of the device for high-resolution measurements of magnetohydrodynamic phenomena including the effects of eddy currents in the nearby conducting wall. The spatial positions of these sensors are calibrated by energizing several large in situ calibration coils in turn, and using measurements of the magnetic fields produced by the various coils to solve for each sensor's position. Since the calibration coils are built near the nominal location of the plasma current centroid, the technique is referred to as an "artificial plasma" calibration. The fitting procedure for the sensor positions is described, and results of the spatial calibration are compared with those based on metrology. The time response of the sensors is compared with the evolution of the artificial plasma current to deduce the eddy current contribution to each signal. This is compared with simulations using the VALEN electromagnetic code, and the modeled copper thickness profiles of the HBT-EP conducting wall are adjusted to better match experimental measurements of the eddy current decay. Finally, the multiple coils of the artificial plasma system are also used to directly calibrate a non-uniformly wound Fourier Rogowski coil on HBT-EP.

  11. Issues in tokamak/stellarator transport and confinement enhancement mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, F.W.

    1990-08-01

    At present, the mechanism for anomalous energy transport in low-{beta} toroidal plasmas -- tokamaks and stellarators -- remains unclear, although transport by turbulent E {times} B velocities associated with nonlinear, fine-scale microinstabilities is a leading candidate. This article discusses basic theoretical concepts of various transport and confinement enhancement mechanisms as well as experimental ramifications which would enable one to distinguish among them and hence identify a dominant transport mechanism. While many of the predictions of fine-scale turbulence are born out by experiment, notable contradictions exist. Projections of ignition margin rest both on the scaling properties of the confinement mechanism and on the criteria for entering enhanced confinement regimes. At present, the greatest uncertainties lie with the basis for scaling confinement enhancement criteria. A series of questions, to be answered by new experimental/theoretical work, is posed to resolve these outstanding contradictions (or refute the fine-scale turbulence model) and to establish confinement enhancement criteria. 73 refs., 4 figs., 5 tabs.

  12. Heating and confinement studies with ECRH in the TCV tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pochelon, A.; Goodman, T.P.; Henderson, M. [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP)] [and others

    1997-06-01

    ECRH experiments have recently started on the TCV tokamak with the use of 1 MW, X2 heating. The ECW system installations is continuing and will eventually provide 3 MW X2 and 1.5 MW X3. The study of the effects of different heating localizations is possible using 1) the mobile mirrors of the launcher, 2) the large vertical room for displacing the plasma in the TCV vessel and 3) the radial displacement of the cyclotron resonance with magnetic field. Initial studies of heating and confinement have concentrated mostly on close-to-circular plasmas to allow the largest variation of beam-plasma geometry and to allow comparison with earlier results on other machines - a necessary first step before investigating more strongly shaped plasmas. For TCV parameters, the nominal field of B = 1.44 T and the frequency of 82.7 GHz of X2 gyrotrons places the resonance position on the high-field side (HFS) of the magnetic axis; while the frequency of 118 GHz of X3 gyrotrons results in a slightly low-field side (LFS) resonance position. (author) 6 figs., 4 refs.

  13. Kinetic modeling of 3D equilibria in a tokamak

    Science.gov (United States)

    Albert, C. G.; Heyn, M. F.; Kasilov, S. V.; Kernbichler, W.; Martitsch, A. F.; Runov, A. M.

    2016-11-01

    External resonant magnetic perturbations (RMPs) can modify the magnetic topology in a tokamak. In this case the magnetic field cannot generally be described by ideal MHD equilibrium equations in the vicinity of resonant magnetic surfaces where parallel and perpendicular relaxation timescales are comparable. Usually, resistive MHD models are used to describe these regions. In the present work, a kinetic model is used for this purpose. Within this model, plasma response, current and charge density are computed with help of a Monte Carlo method, where guiding center orbit equations are solved using a semianalytical geometrical integrator. Besides its higher efficiency in comparison to usual integrators this method is not sensitive to noise in field quantities. The computed charges and currents are used to calculate the electromagnetic field with help of a finite element solver. A preconditioned iterative scheme is applied to search for a self-consistent solution. The discussed method is aimed at the nonlinear kinetic description of RMPs in experiments on Edge Localized Mode (ELM) mitigation by external perturbation coil systems without simplification of the device geometry.

  14. Nonlinear interplay of Alfven instabilities and energetic particles in tokamaks

    CERN Document Server

    Biancalani, A; Cole, M; Di Troia, C; Lauber, Ph; Mishchenko, A; Scott, B; Zonca, F

    2016-01-01

    The confinement of energetic particles (EP) is crucial for an efficient heating of tokamak plasmas. Plasma instabilities such as Alfven Eigenmodes (AE) can redistribute the EP population making the plasma heating less effective, and leading to additional loads on the walls. The nonlinear dynamics of toroidicity induced AE (TAE) is investigated by means of the global gyrokinetic particle-in-cell code ORB5, within the NEMORB project. The nonperturbative nonlinear interplay of TAEs and EP due to the wave-particle nonlinearity is studied. In particular, we focus on the nonlinear modification of the frequency, growth rate and radial structure of the TAE, depending on the evolution of the EP distribution in phase space. For the ITPA benchmark case, we find that the frequency increases when the growth rate decreases, and the mode shrinks radially. This nonlinear evolution is found to be correctly reproduced by means of a quasilinear model, namely a model where the linear effects of the nonlinearly modified EP distri...

  15. Runaway electrons and mitigation studies in MST tokamak plasmas

    Science.gov (United States)

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

    2016-10-01

    Studies of runaway electrons generated in low-density MST tokamak plasmas are being undertaken. The plasmas have Bt resonant magnetic perturbations (RMP's). An m = 3 RMP strongly suppresses the runaway electrons and initial NIMROD modeling shows that this may be due to degradation of flux surfaces. The RMP is produced by a poloidal array of 32 saddle coils at the narrow vertical insulated cut in MST's thick conducting shell, with each RMP having a single m but a broad n spectrum. While a sufficiently strong m = 3 RMP suppresses the runaway electrons, an RMP with m = 1 and comparable amplitude has little effect. The impact of the RMP's on the magnetic topology of these plasmas is being studied with the nonlinear MHD code NIMROD. With an m = 3 RMP, stochasticity is introduced in the outer third of the plasma but no such flux surface degradation is observed with an m = 1 RMP. NIMROD also predicts regularly occurring MHD activity similar to that observed in the experiment. These studies have also been done in q (a) = 2.7 plasmas and analysis and modeling is ongoing. This work supported by USDoE.

  16. Conceptual design of a commercial tokamak hybrid reactor fueling system

    Energy Technology Data Exchange (ETDEWEB)

    Matney, K.D.; Donnert, H.J.; Yang, T.F.

    1979-12-01

    A conceptual design of a fuel injection system for CTHR (Commercial Tokamak Hybrid Reactor) is discussed. Initially, relative merits of the cold-fueling concept are compared with those of the hot-fueling concept; that is, fueling where the electron is below 1 eV is compared with fueling where the electron temperature exceeds 100 eV. It is concluded that cold fueling seems to be somewhat more free of drawbacks than hot fueling. Possible implementation of the cold-fueling concept is exploited via frozen-pellet injection. Several methods of achieving frozen-pellet injection are discussed and the light-gas-gun approach is chosen from these possibilities. A modified version of the ORNL Neutral Gas Shielding Model is used to simulate the pellet injection process. From this simulation, the penetration-depth dependent velocity requirement is determined. Finally, with the velocity requirement known, a gas-pressure requirement for the proposed conceptual design is established. The cryogenic fuel-injection and fuel-handling systems are discussed. A possible way to implement the conceptual device is examined along with the attendant effects on the total system.

  17. Conceptual design of a commercial tokamak hybrid reactor fueling system

    Energy Technology Data Exchange (ETDEWEB)

    Matney, K D; Donnert, H J; Yang, T F

    1979-12-01

    A conceptual design of a fuel injection system for CTHR (Commercial Tokamak Hybrid Reactor) is discussed. Initially, relative merits of the cold-fueling concept are compared with those of the hot-fueling concept; that is, fueling where the electron temperature is below 1 eV is compared with fueling where the electron temperature exceeds 100 eV. It is concluded that cold fueling seems to be somewhat more free of drawbacks than hot fueling. Possible implementation of the cold-fueling concept is exploited via frozen-pellet injection. Several methods of achieving frozen-pellet injection are discussed and the light-gas-gun approach is chosen from these possibilities. A modified version of the ORNL Neutral Gas Shielding Model is used to simulate the pellet injection process. From this simulation, the penetration-depth dependent velocity requirement is determined. Finally, with the velocity requirement known, a gas-pressure requirement for the proposed conceptual design is established. The cryogenic fuel-injection and fuel-handling systems are discussed. A possible way to implement the conceptual device is examined along with the attendant effects on the total system.

  18. Sawtooth Activity in Ohmically Heated Plasma on HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Sawtooth activity on HT-7 tokamak has been investigated experimentally mainly by using soft x-ray diode array and magnetic probes. Their behaviors and occurrences are correlatedclosely to the discharge conditions: the electron density Ne, the electron temperature Te, the safetyfactor qa on plasma boundary and wall condition etc. When central line-averaged electron densityNe(0) is over 2.0×1013cm-3, major sawtooth activity emerges with a period of up to 6.5 ms and afluctuation amplitude of up to 2~30 % of SXR radiation signal. In some cases such as the safetyfactor between 4.2~4.7 and Zeff=3.0~6.0, a monster sawtooth activity often emerges withoutapparent deterioration of plasma confinement and without major disruption. During these events,abundant MHD phenomena are observed including partial sawtooth oscillations. In this paper, theobserved sawtooth behaviors and their dependence on the and their dependence density Ne andwall condition in ohmically heated plasma are introduced, the results are discussed and presented.

  19. Theory of nonaxisymmetric vertical displacement events in tokamaks

    Science.gov (United States)

    Fitzpatrick, R.

    2011-05-01

    A semi-analytic sharp-boundary model of a nonaxisymmetric vertical displacement event (VDE) in a large aspect-ratio, high-beta (i.e. β ~ epsilon), vertically elongated tokamak plasma is developed. The model is used to simulate nonaxisymmetric VDEs with a wide range of different plasma equilibrium and vacuum vessel parameters. These simulations yield poloidal halo current fractions and toroidal peaking factors whose magnitudes are similar to those seen in experiments, and also reproduce the characteristic inverse scaling between the halo current fraction and the toroidal peaking factor. Moreover, the peak poloidal halo current density in the vacuum vessel is found to correlate strongly with the reciprocal of the minimum edge safety factor attained during the VDE. In addition, under certain circumstances, the ratio of the net sideways force acting on the vacuum vessel to the net vertical force is observed to approach unity. Finally, the peak vertical force per unit area acting on the vessel is found to have a strong correlation with the equilibrium toroidal plasma current at the start of the VDE, but is also found to increase with increasing vacuum vessel resistivity relative to the scrape-off layer plasma.

  20. Initial Plasma Startup Test on SUNIST Spherical Tokamak

    Institute of Scientific and Technical Information of China (English)

    Wang Ying(王莹); Zeng Li(曾立); He Yexi(何也熙); SUMST Team

    2003-01-01

    The goal of the Sino-United Spherical Tokamak (SUNIST) at Tsinghua University is to extend the understanding of toroidal plasma physics at a low aspect ratio (R/a ≈ 1.3) and to demonstrate a maintainable target plasma by non-inductive startup. The SUNIST device is designed to operate with up to 13 kA of ohmic heating field current, and to 0.15 T of toroidal field at 10 kA of discharge current. All of the poloidal fields can provide 30 mVs of Volt-seconds transformer. Experimental results of plasma startup show that SUNIST has remarkable characteristics of high ramp rate (dIp/dt ≈ 50 MA/s ), high normalized current IN of about 2.8 (IN = Ip/aBT),and high-efficiency (Ip/IROD ≈ 0.4) production of plasma current while operating at a low toroidal field. Major disruption phenomena have not been observed from magnetic diagnostics of all testing shots. Initial discharges with 52 kA of plasma current (exceeding the designed value of 50 kA),2 ms of pulse length and 50 MA/s of ramp rate have been achieved easily with pre-ionized filament.

  1. Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Green, David L [ORNL; Jaeger, E. F. [XCEL; Berry, Lee A [ORNL; Chen, Guangye [ORNL; Ryan, Philip Michael [ORNL; Canik, John [ORNL

    2011-01-01

    Observations of improved radio frequency (RF) heating efficiency in high-confinement (H-) mode plasmas on the National Spherical Tokamak Experiment (NSTX) are investigated by whole-device linear simulation. We present the first full-wave simulation to couple kinetic physics of the well confined core plasma to the poorly confined scrape-off plasma. The new simulation is used to scan the launched fast-wave spectrum and examine the steady-state electric wave field structure for experimental scenarios corresponding to both reduced, and improved RF heating efficiency. We find that launching toroidal wave-numbers that required for fast-wave propagation excites large amplitude (kVm 1 ) coaxial standing modes in the wave electric field between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggest these modes are a probable cause of degraded heating efficiency. Also, the H-mode density pedestal and fast-wave cutoff within the confined plasma allow for the excitation of whispering gallery type eigenmodes localised to the plasma edge.

  2. Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak.

    Science.gov (United States)

    Qu, Hao; Zhang, Tao; Han, Xiang; Wen, Fei; Zhang, Shoubiao; Kong, Defeng; Wang, Yumin; Gao, Yu; Huang, Canbin; Cai, Jianqing; Gao, Xiang

    2015-08-01

    An X-mode polarized V band (50 GHz-75 GHz) radial and poloidal correlation reflectometry is designed and installed on Experimental Advanced Superconducting Tokamak (EAST). Two frequency synthesizers (12 GHz-19 GHz) are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together for launching through one single pyramidal antenna. Two poloidally separated antennae are installed to receive the reflected waves from plasma. This reflectometry system can be used for radial and poloidal correlation measurement of the electron density fluctuation. In ohmically heated plasma, the radial correlation length is about 1.5 cm measured by the system. The poloidal correlation analysis provides a means to estimate the fluctuation velocity perpendicular to the main magnetic field. In the present paper, the distance between two poloidal probing points is calculated with ray-tracing code and the propagation time is deduced from cross-phase spectrum. Fluctuation velocity perpendicular to the main magnetic field in the core of ohmically heated plasma is about from -1 km/s to -3 km/s.

  3. Design of the Cryostat for HT—7U superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

    郁杰; 武松涛; 等

    2002-01-01

    The cryostat of HT-7U tokamak is a large vacuum vessel surrounding the entire basic machine with a cylindrical shell,a dished top and a flat bottom.The main function of HT-7U cryostat is to provide a thermal barrier between an ambient temperature test hall and a liquid helium-cooled superconducting magnet.The loads applied to the cryostat are from sources of vacuum pressure,dead weight,seismic events and electromagnetic forces originated by eddy currents.It also provides feed-through penetrations for all the conecting elements inside and outside the cryostat.The main material selected for the cryostat is stainless steel 304L.The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out by using a finite element code.Stress analysis results show that the maximum stress intensity was below the allowable value.In this paper,the structural analyses and design of HT-7U cryostat are emphasized.

  4. THz time-domain spectroscopy for tokamak plasma diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Causa, F.; Zerbini, M.; Buratti, P.; Gabellieri, L.; Pacella, D.; Romano, A.; Tuccillo, A. A.; Tudisco, O. [ASSOCIAZIONE EURATOM ENEA sulla Fusione, C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Johnston, M. [Clarendon Laboratory, Department of Physics, University of Oxford, Parks Road, Oxford OX1 3PU (United Kingdom); Doria, A.; Gallerano, G. P.; Giovenale, E. [ENEA C.R. Frascati UTAPRAD, via E. Fermi 45, 00044 Frascati (Roma) (Italy)

    2014-08-21

    The technology is now becoming mature for diagnostics using large portions of the electromagnetic spectrum simultaneously, in the form of THz pulses. THz radiation-based techniques have become feasible for a variety of applications, e.g., spectroscopy, imaging for security, medicine and pharmaceutical industry. In particular, time-domain spectroscopy (TDS) is now being used also for plasma diagnostics in various fields of application. This technique is promising also for plasmas for fusion applications, where plasma characteristics are non-uniform and/or evolve during the discharge This is because THz pulses produced with femtosecond mode-locked lasers conveniently span the spectrum above and below the plasma frequency and, thus, can be used as very sensitive and versatile probes of widely varying plasma parameters. The short pulse duration permits time resolving plasma characteristics while the large frequency span permits a large dynamic range. The focus of this work is to present preliminary experimental and simulation results demonstrating that THz TDS can be realistically adapted as a versatile tokamak plasma diagnostic technique.

  5. Control of bootstrap current in the pedestal region of tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Shaing, K. C. [Institute for Space and Plasma Sciences, National Cheng Kung University, Tainan City 70101, Taiwan (China); Department of Engineering Physics, University of Wisconsin, Madison, Wisconsin 53796 (United States); Lai, A. L. [Institute for Space and Plasma Sciences, National Cheng Kung University, Tainan City 70101, Taiwan (China)

    2013-12-15

    The high confinement mode (H-mode) plasmas in the pedestal region of tokamaks are characterized by steep gradient of the radial electric field, and sonic poloidal U{sub p,m} flow that consists of poloidal components of the E×B flow and the plasma flow velocity that is parallel to the magnetic field B. Here, E is the electric field. The bootstrap current that is important for the equilibrium, and stability of the pedestal of H-mode plasmas is shown to have an expression different from that in the conventional theory. In the limit where ‖U{sub p,m}‖≫ 1, the bootstrap current is driven by the electron temperature gradient and inductive electric field fundamentally different from that in the conventional theory. The bootstrap current in the pedestal region can be controlled through manipulating U{sub p,m} and the gradient of the radial electric. This, in turn, can control plasma stability such as edge-localized modes. Quantitative evaluations of various coefficients are shown to illustrate that the bootstrap current remains finite when ‖U{sub p,m}‖ approaches infinite and to provide indications how to control the bootstrap current. Approximate analytic expressions for viscous coefficients that join results in the banana and plateau-Pfirsch-Schluter regimes are presented to facilitate bootstrap and neoclassical transport simulations in the pedestal region.

  6. Real Time Equilibrium Reconstruction Algorithm in EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    王华忠; 罗家融; 黄勤超

    2004-01-01

    The EAST (HT-7U) superconducting tokamak is a national project of China on fusion research, with a capability of long-pulse (~ 1000 s) operation. In order to realize a longduration steady-state operation of EAST, some significant capability of real-time control is required. It would be very crucial to obtain the current profile parameters and the plasma shapes in real time by a flexible control system. As those discharge parameters cannot be directly measured,so a current profile consistent with the magnetohydrodynamic equilibrium should be evaluated from external magnetic measurements, based on a linearized iterative least square method, which can meet the requirements of the measurements. The arithmetic that the EFIT (equilibrium fitting code) is used for reference will be given in this paper and the computational efforts are reduced by parametrizing the current profile linearly in terms of a number of physical parameters.In order to introduce this reconstruction algorithm clearly, the main hardware design will be listed also.

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

    Science.gov (United States)

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

    2014-08-01

    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.

  8. Analysis of Tangential Camera Views of Tokamak Plasmas

    Science.gov (United States)

    Meyer, W. H.; Fenstermacher, M. E.; Groth, M.

    2014-10-01

    Commercial digital video cameras are increasingly being employed as sensor and digitizer for fusion plasma diagnostics. Here we describe some of our recent work to significantly improve tomographic analysis of tangential viewing camera data in toroidal geometry. Fiducial images obtained during vents are used to produce direct linear transformation matrices for each diagnostic camera and determine an accurate camera position and viewport. A tokamak solid model for rendering is used between vents to check camera alignment or when hardware fiducial images are unavailable. The rendering provides camera pixel distance of integration, vessel impact position, and angle of incidence. Reflection characterization is also performed using this solid model. After determining the camera geometry parameters, a response matrix is calculated for toroidally symmetric solution grids. We apply standard sparse linear solvers to the back projection problem for filtered visible light cameras. Forward projections are utilized for theory code validation and visualization. Poloidal IR profiles are extracted from tangential viewing IR camera data using surface parameters determined from solid model rendering. We present examples of each of these applications. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

  9. Ex-vessel remote maintenance for the Compact Ignition Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist in removing and repairing such components as diagnostic modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the plasma chamber includes a bridge-mounted manipulator system for test cell operations, decontamination (decon) equipment, hot cell equipment, and solid-radiation-waste-handling equipment. Wherever possible, the project will use commercially available equipment. Several areas of the maintenance system design were addressed in fiscal year (FY) 1987, including conceptual designs of manipulator systems, the start of a remote equipment research and development (RandD) program, and definition of the hot cell, decon, and equipment repair facility requirements. R and D work included preliminary demonstrations of remote handling operations on full-size, partial mock-ups of the CIT machine at the Oak Ridge National Laboratory (ORNL) Remote Operations and Maintenance Development (ROMD) Facility. 1 ref., 6 figs.

  10. Optical Design of ECEI Diagnostic System for HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    Wang Jun(王俊); Wen Yizhi(闻一之); Yu Changxuan(俞昌旋); Wan Baonian(万宝年); N.C. Luhmann; Wang Jian; Xia Z. G.

    2004-01-01

    Electron cyclotron emission imaging system in the frequency range of 95 GHz ~125 GHz is going to be constructed for a two-dimensional diagnosis of the electron temperature profiles and fluctuations on the HT-7 Tokamak. The optical design for the ECEI diagnostic system is completed. Because of the superconducting technology used in HT-7, the vacuum chamber is rather thick (630 mm), the height of the horizontal windows is limited (maximum 450 mm), which constrains greatly the ECE imaging Gaussian beam that passing through the windows. We here comes to make a design compromise between the number of the beams that can pass through the windows and the spatial resolution (around 1.1 cm). We also find that due to the field curvature of the optical system, the gaussian beams of edge channels are always overlapped. To flatten the field curvature, it is needed to insert a concave made of a material with a low refractive index (compared with the one used in the convex). But the suitable material has not been available so far, therefore the deterioration of the resolution in some channels (e.g. the edge channels) is acceptable.

  11. Modelling of electron transport and of sawtooth activity in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Angioni, C

    2001-10-01

    Transport phenomena in tokamak plasmas strongly limit the particle and energy confinement and represent a crucial obstacle to controlled thermonuclear fusion. Within the vast framework of transport studies, three topics have been tackled in the present thesis: first, the computation of neoclassical transport coefficients for general axisymmetric equilibria and arbitrary collisionality regime; second, the analysis of the electron temperature behaviour and transport modelling of plasma discharges in the Tokamak a configuration Variable (TCV); third, the modelling and simulation of the sawtooth activity with different plasma heating conditions. The work dedicated to neoclassical theory has been undertaken in order to first analytically identify a set of equations suited for implementation in existing Fokker-Planck codes. Modifications of these codes enabled us to compute the neoclassical transport coefficients considering different realistic magnetic equilibrium configurations and covering a large range of variation of three key parameters: aspect ratio, collisionality, and effective charge number. A comparison of the numerical results with an analytical limit has permitted the identification of two expressions for the trapped particle fraction, capable of encapsulating the geometrical effects and thus enabling each transport coefficient to be fitted with a single analytical function. This has allowed us to provide simple analytical formulae for all the neoclassical transport coefficients valid for arbitrary aspect ratio and collisionality in general realistic geometry. This work is particularly useful for a correct evaluation of the neoclassical contribution in tokamak scenarios with large bootstrap cur- rent fraction, or improved confinement regimes with low anomalous transport and for the determination of the plasma current density profile, since the plasma conductivity is usually assumed neoclassical. These results have been included in the plasma transport code

  12. Litization of FTU tokamak vacuum vessel by using a Li limiter

    Science.gov (United States)

    Mazzitelli, Giuseppe; Apicella, Maria Laura; Lazarev, Vladimir; Azizov, E. A.; Mirnov, S. V.; Petrov, Vladimir; Vertkov, Alexei; Evtikhin, V. A.; Lyublinski, I. E.

    2003-10-01

    The idea to use lithium, a low Z metallic material, for the first wall protection of the tokamak-reactor was known for a rather long time. However practical application of lithium on experimental installations began rather recently. Litization - the coating of tokamak vacuum vessel walls by a thin lithium film is one of methods for protection of plasma from high Z impurities fluxes. A possible method for litization of the vacuum walls is by sputtering and evaporation of a Li limiter . This method utilizes a new concept of limiter on the basis of capillary-porous structure (CPS), having a sufficient mechanical stability (Li confinement). Furthermore the proposed experiment aims to study the liquid Li erosion, Li accumulation and distribution in plasma, thermal load reduction on the limiter due to Li radiation losses in a medium size tokamak as FTU.

  13. An H{sup {infinity}} system identification algorithm applied to tokamak modelling

    Energy Technology Data Exchange (ETDEWEB)

    Coutlis, A.; Limebeer, D.J.N.; Wainwright, J. [Centre for Process Systems Eng. and Dept. of Electrical Eng., Imperial College, London (United Kingdom); Lister, J.B.; Vyas, P.; Ward, D.J. [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP)

    1997-08-01

    In this paper we describe the application of MISO System Identification to Tokamak simulations and machines. The work is motivated by the desire to create linear models for the design of modern controllers. The method described in this paper is a worst-case identification technique, in that it aims to minimise the H{sup {infinity}} error between the identified model and the plant. Such a model is particularly suited for robust controller design. The method is fully detailed from the design of identification experiments through to the creation of a low-order model from a combination of Hankel model reduction and Chebycheff approximation. We show results from the application of this method to a powerful Tokamak Simulation Code (TSC) and discuss results on the TCV Tokamak in Lausanne. (author) 2 figs., 11 refs.

  14. Effect of magnetic perturbations on the 3D MHD self-organization of shaped tokamak plasmas

    CERN Document Server

    Bonfiglio, D; Veranda, M; Chacón, L; Escande, D F

    2016-01-01

    The effect of magnetic perturbations (MPs) on the helical self-organization of shaped tokamak plasmas is discussed in the framework of the nonlinear 3D MHD model. Numerical simulations performed in toroidal geometry with the \\textsc{pixie3d} code [L. Chac\\'on, Phys. Plasmas {\\bf 15}, 056103 (2008)] show that $n=1$ MPs significantly affect the spontaneous quasi-periodic sawtoothing activity of such plasmas. In particular, the mitigation of sawtooth oscillations is induced by $m/n=1/1$ and $2/1$ MPs. These numerical findings provide a confirmation of previous circular tokamak simulations, and are in agreement with tokamak experiments in the RFX-mod and DIII-D devices. Sawtooth mitigation via MPs has also been observed in reversed-field pinch simulations and experiments. The effect of MPs on the stochastization of the edge magnetic field is also discussed.

  15. The Experiment of Modulated Toroidal Current on HT-7 and HT-6M Tokamak

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The Experiments of Modulated Toroidal Current were done on the HT-6M tokamakand HT-7 superconducting tokamak. The toroidal current was modulated by programming theOhmic heating field. Modulation of the plasma current has been used successfully to suppressMHD activity in discharges near the density limit where large MHD m = 2 tearing modes weresuppressed by sufficiently large plasma current oscillations. The improved Ohmic confinementphase was observed during modulating toroidal current (MTC) on the Hefei Tokamak-6M (HT-6M) and Hefei superconducting Tokamak-7 (HT-7). A toroidal frequency-modulated current,induced by a modulated loop voltage, was added on the plasma equilibrium current. The ratio ofA.C. amplitude of plasma current to the main plasma current △Ip/Ip is about 12% ~ 30%. Thedifferent formats of the frequency-modulated toroidal current were compared.

  16. High performance discharges in the Lithium Tokamak eXperiment with liquid lithium wallsa)

    Science.gov (United States)

    Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G.; Capece, A.; Koel, B.; Roszell, J.; Biewer, T. M.; Gray, T. K.; Kubota, S.; Beiersdorfer, P.; Widmann, K.; Tritz, K.

    2015-05-01

    The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

  17. A discrete adaptive near-time optimum control for the plasma vertical position in a Tokamak

    CERN Document Server

    Scibile, L

    2001-01-01

    A nonlinear controller for the plasma vertical position in a Tokamak, based on a discrete-time adaptive near time optimum control algorithm (DANTOC) is designed to stabilize the system and to maximize the state-space region over which stability can be guaranteed. The controller is also robust to the edge localized modes (ELMs) and the 600 Hz noise from the thyristor power supplies that are the primary source of disturbances and measurement noise. The controller is tested in simulation for the JET Tokamak and the results confirm its efficacy in controlling the vertical position for different plasma configurations. The controller is also tested experimentally on a real Tokamak, COMPASS-D, and the results demonstrate the improvement with respect to a simple linear PD controller in the presence of disturbances and measurement noise. The emphasis of the is on the development of the design methodology. (38 refs).

  18. Toothbrush probe for instantaneous measurement of radial profile in tokamak boundary plasma

    Energy Technology Data Exchange (ETDEWEB)

    Uehara, Kazuya; Sengoku, Seio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Amemiya, Hiroshi

    1997-04-01

    A new probe for the instantaneous measurement of radial profiles of the boundary scrape-off layer (SOL) plasma has been developed in a tokamak. Five asymmetric double-probe chips are aligned in parallel to a strong magnetic field in the boundary plasma in a tokamak. This probe is named the `toothbrush probe` and can measure the ion temperature as well as the electron temperature and the plasma density in the SOL plasma within only one tokamak plasma shot. First, only one asymmetric probe is mounted on the divertor plate and it is tried to determine the ion temperature. Then, a manufactured toothbrush probe is mounted in the SOL plasma and the radial plasma profiles are simultaneously obtained. Data on the e-folding length of the plasma profile obtained by the toothbrush probe can determine the information on the transport properties such as the diffusion coefficient and the thermal conductivity of electrons and ions. (author)

  19. Transmutation of nuclear waste with a low-aspect-ratio Tokamak neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Moon, Se Youn [Chonbuk National University, Jeonju (Korea, Republic of)

    2014-10-15

    The transmutation characteristics of transuranics (TRUs) in a transmutation reactor based on a LAR (Low-aspect-ratio) tokamak as a neutron source are investigated. The optimum radial build of a transmutation reactor is found by using a coupled analysis of the tokamak systems and the neutron transport. The dependences of the transmutation characteristics on the aspect ratio A in the range of 1.5 to 2.5 and on the fusion power in the range of 150 to 500 MW are investigated. An equilibrium fuel cycle is developed for effective transmutation, and show that with one unit of the transmutation reactor based on the LAR tokamak producing fusion power in the range of a few hundred MWs, up to 3 PWRs (1.0 GWe capacity) can be supported with a burn-up fraction larger than 50%.

  20. Transmutation of nuclear waste with a low-aspect-ratio tokamak neutron source

    Science.gov (United States)

    Hong, Bong Guen; Moon, Se Youn

    2014-10-01

    The transmutation characteristics of transuranics (TRUs) in a transmutation reactor based on a LAR (Low-aspect-ratio) tokamak as a neutron source are investigated. The optimum radial build of a transmutation reactor is found by using a coupled analysis of the tokamak systems and the neutron transport. The dependences of the transmutation characteristics on the aspect ratio A in the range of 1.5 to 2.5 and on the fusion power in the range of 150 to 500 MW are investigated. An equilibrium fuel cycle is developed for effective transmutation, and show that with one unit of the transmutation reactor based on the LAR tokamak producing fusion power in the range of a few hundred MWs, up to 3 PWRs (1.0 GWe capacity) can be supported with a burn-up fraction larger than 50%.

  1. Macroscopic erosion of divertor and first wall armour in future tokamaks

    Science.gov (United States)

    Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

    2002-12-01

    Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

  2. Magnetized plasma flow injection into tokamak and high-beta compact torus plasmas

    Science.gov (United States)

    Matsunaga, Hiroyuki; Komoriya, Yuuki; Tazawa, Hiroyasu; Asai, Tomohiko; Takahashi, Tsutomu; Steinhauer, Loren; Itagaki, Hirotomo; Onchi, Takumi; Hirose, Akira

    2010-11-01

    As an application of a magnetized coaxial plasma gun (MCPG), magnetic helicity injection via injection of a highly elongated compact torus (magnetized plasma flow: MPF) has been conducted on both tokamak and field-reversed configuration (FRC) plasmas. The injected plasmoid has significant amounts of helicity and particle contents and has been proposed as a fueling and a current drive method for various torus systems. In the FRC, MPF is expected to generate partially spherical tokamak like FRC equilibrium by injecting a significant amount of magnetic helicity. As a circumstantial evidence of the modified equilibrium, suppressed rotational instability with toroidal mode number n = 2. MPF injection experiments have also been applied to the STOR-M tokamak as a start-up and current drive method. Differences in the responses of targets especially relation with beta value and the self-organization feature will be studied.

  3. Translational symmetry of high order tokamak flux surface shaping in gyrokinetics

    CERN Document Server

    Ball, Justin; Barnes, Michael

    2015-01-01

    A particular translational symmetry of the local nonlinear $\\delta f$ gyrokinetic model is demonstrated analytically and verified numerically. This symmetry shows that poloidally translating all the flux surface shaping effects with large poloidal mode number by a single tilt angle has an exponentially small effect on the transport properties of a tokamak. This is shown using a generalization of the Miller local equilibrium model to specify an arbitrary flux surface geometry. With this geometry specification we find that, when performing an expansion in large flux surface shaping mode number, the governing equations of gyrokinetics are symmetric in the poloidal translation of the high order shaping effects. This allows us to take the fluxes from a single configuration and calculate the fluxes in any configuration that can be produced by translating the large mode number shaping effects. This creates a distinction between tokamaks with mirror symmetric flux surfaces and tokamaks without mirror symmetry, which ...

  4. Real-time identification of the current density profile in the JET Tokamak: method and validation

    CERN Document Server

    Mazon, Didier; Boulbe, Cédric; Faugeras, Blaise; Boboc, A; Brix, M; De Vries, P; Sharapov, S; Zabeo, L

    2009-01-01

    The real-time reconstruction of the plasma magnetic equilibrium in a Tokamak is a key point to access high performance regimes. Indeed, the shape of the plasma current density profile is a direct output of the reconstruction and has a leading effect for reaching a steady-state high performance regime of operation. In this paper we present the methodology followed to identify numerically the plasma current density in a Tokamak and its equilibrium. In order to meet the real-time requirements a C++ software has been developed using the combination of a finite element method, a nonlinear fixed point algorithm associated to a least square optimization procedure. The experimental measurements that enable the identification are the magnetics on the vacuum vessel, the interferometric and polarimetric measurements on several chords and the motional Stark effect. Details are given about the validation of the reconstruction on the JET tokamak, either by comparison with ?off-line' equilibrium codes or real time software ...

  5. Spectroscopic diagnostics of superthermal electrons with high-number harmonic EC radiation in tokamak reactor plasmas

    Directory of Open Access Journals (Sweden)

    Minashin P.V.

    2015-01-01

    Full Text Available A method of spectroscopic diagnostics of the average perpendicular-to-magnetic-field momentum of the superthermal component of the electron velocity distribution (EVD, based on the high-number-harmonic electron cyclotron (EC radiation, is suggested for nuclear fusion-reactor plasmas under condition of a strong auxiliary heating (e.g. in tokamak DEMO, a next step after tokamak ITER. The method is based on solving an inverse problem for reconstruction of the EVD in parallel and perpendicular-to-magnetic-field components of electron momentum at high and moderate energies responsible for the emission of the high-number-harmonic EC radiation.

  6. Impurity Measured by VUV Spectrometer and OMA on HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    崔正英; 王全明; 董贾福; 孙平; 卢杰; 李伟; 王恩耀

    2004-01-01

    Impurity is one of the key issues on a great impact to the quality of tokamak plasma.HL-2A is the first divertor tokamak in China. In this paper the experimental results are presented on impurity through the line emission measurement in the campaign in 2003 under the limiter and divertor configurations. The low-Z impurities such as carbon and oxygen are the most important components in the plasma, but their content are not so high to affect the discharge quality. The high-Z impurities such as copper and ferrum are not essential. The emission intensity of impurity is clearly decreased during the divertor configuration formed.

  7. Poloidal beta and internal inductance measurement on HT-7 superconducting tokamak.

    Science.gov (United States)

    Shen, B; Sun, Y W; Wan, B N; Qian, J P

    2007-09-01

    Poloidal beta beta(theta) and internal inductance l(i) measurements are very important for tokamak operation. Much more plasma parameters can be inferred from the two parameters, such as the plasma energy confinement time, the plasma toroidal current profile, and magnetohydrodynamics instability. Using diamagnetic and compensation loop, combining with poloidal magnetic probe array signals, poloidal beta beta(theta) and internal inductance l(i) are measured. In this article, the measurement system and arithmetic are introduced. Different experimental results are given in different plasma discharges on HT-7 superconducting tokamak.

  8. 3D MHD VDE and disruptions simulations of tokamaks plasmas including some ITER scenarios

    Science.gov (United States)

    Paccagnella, R.; Strauss, H. R.; Breslau, J.

    2009-03-01

    Tokamaks vertical displacement events (VDEs) and disruptions simulations in toroidal geometry by means of a single fluid visco-resistive magneto-hydro-dynamic (MHD) model are presented in this paper. The plasma model is completed with the presence of a 2D wall with finite resistivity which allows the study of the relatively slowly growing magnetic perturbation, the resistive wall mode (RWM), which is, in this paper, the main drive of the disruption evolution. Amplitudes and asymmetries of the halo currents pattern at the wall are also calculated and comparisons with tokamak experimental databases and predictions for ITER are given.

  9. Tangential and Vertical Compact Torus Injection Experiments on the STOR-M Tokamak

    Institute of Scientific and Technical Information of China (English)

    Xiao Chijin; Liu D.; S. Livingstone; A. K. Singh; E. Zhang; A. Hirose

    2005-01-01

    This paper describes the setup and results of compact torus (CT) injection experiments on the STOR-M tokamak. Tangential CT injection into STOR-M induced H-mode-like phenomena including doubling the electron density, reduction in the Ha radiation level, suppression of the floating potential fluctuations, suppression of the m = 2 Mirnov oscillations, and increase in the global energy confinement time. Experimental setup, bench-test results, and some preliminary injection data for vertical CT injection experiments on STOR-M will be shown. In addition, numerical simulations of the CT trajectories in tokamak discharges for both tangential and vertical injection geometries will be discussed.

  10. Remote servicing considerations for near term tokamak power reactors (TNS). Final summary

    Energy Technology Data Exchange (ETDEWEB)

    Spampinato, P.T.

    1977-01-01

    Next generation Tokamaks require special consideration for remote servicing. Three major problems are highlighted: (1) movement of heavy components, (2) remote connection/disconnection of joints, and (3) remote cutting, welding, and leak detection. The first problem is assumed to be handled with existing expertise and is not considered. The remaining problems are thought to be minimized by considering two engineering departures from conventional tokamak design; locating the field shaping coils outside of the toroidal coils and enclosing the total device within an evacuated reactor cell. Five topics under this vacuum building concept are discussed: incremental cost, vacuum pumping, tritium containment, activation topology, and first year operations.

  11. Disruption mitigation using laser ablation of high-Z impurities in HL-1M tokamak

    Institute of Scientific and Technical Information of China (English)

    Zheng Yong-Zhen; Feng Xing-Ya; Guo Gan-Cheng; Xu De-Ming; Zheng Yin-Jia

    2006-01-01

    A preliminary experiment triggering a plasma current quench by laser ablation of high-Z impurities has been performed in the HL-1M tokamak. The injection of impurities with higher electric charges into tokamak plasmas can increase the radiation cooling of the plasma. Resistive, highly radiating plasma formed prior to the thermal quench can dissipate both the thermal and magnetic energies, which is possibly a simple and potential approach to reducing significantly the plasma thermal energy and magnetic energy before a disruption thereby a safe plasma termination is obtained.

  12. Effect of energy and momentum conservation on fluid resonances for resonant magnetic perturbations in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Leitner, Peter; Heyn, Martin F.; Kernbichler, Winfried [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, TU Graz, Petersgasse 16, A-8010 Graz (Austria); Ivanov, Ivan B. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, TU Graz, Petersgasse 16, A-8010 Graz (Austria); St. Petersburg State University, Institute of Physics, Ulyanovskaya 1, Petrodvoretz 198504 (Russian Federation); Petersburg Nuclear Physics Institute, 188300 Gatchina, Leningrad Oblast (Russian Federation); Kasilov, Sergei V. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, TU Graz, Petersgasse 16, A-8010 Graz (Austria); Institute of Plasma Physics, National Science Center “Kharkov Institute of Physics and Technology,” Ul. Akademicheskaya 1, 61108 Kharkov (Ukraine)

    2014-06-15

    In this paper, the impact of momentum and energy conservation of the collision operator in the kinetic description for Resonant Magnetic Perturbations (RMPs) in a tokamak is studied. The particle conserving differential collision operator of Ornstein-Uhlenbeck type is supplemented with integral parts such that energy and momentum are conserved. The application to RMP penetration in a tokamak shows that energy conservation in the electron collision operator is important for the quantitative description of plasma shielding effects at the resonant surface. On the other hand, momentum conservation in the ion collision operator does not significantly change the results.

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

    Science.gov (United States)

    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

  14. Preliminary Design of Control Network for HT-7U Tokamak Cryogenic System

    Institute of Scientific and Technical Information of China (English)

    Jin Yibin(金毅彬); Zhuang Ming(庄明); Bai Hongyu(白宏宇)

    2003-01-01

    In the course of the cryoplant modernization, a control network will be set up in order to facilitate the control, the supervision, the centralized data acquisition and the alarm handling of the cryogenic system for HT-7U tokamak. The paper introduces the preliminary design of control network based on the Controller Link Network for HT-7U tokamak cryogenic system. The multi-layer structure mentioned in the paper is the mainstream of automatic control.The control philosophy, the structure of the network and the components for control are also presented.

  15. Determination of confinement efficiency in tokamaks based on current independent flux loops technique

    Science.gov (United States)

    Salar Elahi, A.; Ghoranneviss, M.

    In this contribution we presented a current independent approximation of the combination of poloidal beta and internal inductance (confinement efficiency) only based on poloidal flux loops measurement in IR-T1 tokamak. The main advantage of this technique is that it based only on the one diagnostic (only flux loops and not need to plasma current measurement). Based on this method, two flux loops were designed, constructed, and installed on outer surface of the IR-T1 tokamak chamber and then the Shafranov parameter was measured from them. Also the result of this technique was compared with conventional magnetic probes technique and found in good agreement with each other.

  16. Note: Upgrade of electron cyclotron emission imaging system and preliminary results on HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, M., E-mail: jiangm@swip.ac.cn; Shi, Z. B.; Zhong, W. L.; Chen, W.; Liu, Z. T.; Ding, X. T.; Yang, Q. W.; Zhang, B. Y.; Shi, P. W.; Liu, Y.; Fu, B. Z.; Xu, Y. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Domier, C. W.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California, Davis, California 95616 (United States); Yang, Z. C. [School of Physics and Chemistry, Xihua University, Chengdu 610039 (China)

    2015-07-15

    The electron cyclotron emission imaging system on the HL-2A tokamak has been upgraded to 24 (poloidally) × 16 (radially) channels based on the previous 24 × 8 array. The measurement region can be flexibly shifted due to the independence of the two local oscillator sources, and the field of view can be adjusted easily by changing the position of the zoom lenses. The temporal resolution is about 2.5 μs and the achievable spatial resolution is 1 cm. After laboratory calibration, it was installed on HL-2A tokamak in 2014, and the local 2D mode structures of MHD activities were obtained for the first time.

  17. Distributed digital real-time control system for TCV tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Le, H.B. [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association EURATOM-Confédération Suisse, CH-1015 Lausanne (Switzerland); Felici, F. [Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Paley, J.I.; Duval, B.P.; Moret, J.-M.; Coda, S.; Sauter, O.; Fasel, D.; Marmillod, P. [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association EURATOM-Confédération Suisse, CH-1015 Lausanne (Switzerland)

    2014-03-15

    Highlights: • A new distributed digital control system for the TCV tokamak has been commissioned. • Data is shared in real-time between all nodes using the reflective memory. • The customised Linux OS allows achieving deterministic and low latency behaviour. • The control algorithm design in Simulink together with the automatic code generation using Embedded Coder allow rapid algorithm development. • Controllers designed outside the TCV environment can be ported easily. • The previous control system functions have been emulated and improved. • New capabilities include MHD control, profile control, equilibrium reconstruction. - Abstract: A new digital feedback control system (named the SCD “Système de Contrôle Distribué”) has been developed, integrated and used successfully to control TCV (Tokamak à Configuration Variable) plasmas. The system is designed to be modular, distributed, and scalable, accommodating hundreds of diagnostic inputs and actuator outputs. With many more inputs and outputs available than previously possible, it offers the possibility to design advanced control algorithms with better knowledge of the plasma state and to coherently control all TCV actuators, including poloidal field (PF) coils, gas valves, the gyrotron powers and launcher angles of the electron cyclotron heating and current drive system (ECRH/ECCD) together with diagnostic triggering signals. The system consists of multiple nodes; each is a customised Linux desktop or embedded PC which may have local ADC and DAC cards. Each node is also connected to a memory network (reflective memory) providing a reliable, deterministic method of sharing memory between all nodes. Control algorithms are programmed as block diagrams in Matlab-Simulink providing a powerful environment for modelling and control design. The C code is generated automatically from the Simulink block diagram and compiled, with the Simulink Embedded Coder (SEC, formerly Real-Time Workshop Embedded

  18. Characterization and scaling of the tokamak edge transport barrier

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Philip Adrian

    2012-04-24

    The high confinement regime (H-mode) in a tokamak plasma displays a remarkable edge region. On a small spatial scale of 1-2 cm the properties of the plasma change significantly. Certain parameters vary 1-2 orders of magnitude in this region, called the pedestal. Currently, there is no complete understanding of how the pedestal forms or how it is sustained. The goal of this thesis is to contribute to the theoretical understanding of the pedestal and provide scalings towards larger machines, like ITER and DEMO. A pedestal database was built with data from different tokamaks: ASDEX Upgrade, DIIID and JET. The pedestal was characterized with the same method for all three machines. This gives the maximum value, gradient and width of the pedestal in n{sub e}, T{sub e} and T{sub i}. These quantities were analysed along with quantities derived from them, such as the pressure or the confinement time. For this purpose two parameter sets were used: normalized parameters (pressure {beta}, time {nu}{sub *}, length {rho}{sub *}, shape f{sub q}) and machine parameters (size a, magnetic field B{sub t}, plasma current I{sub p}, heating P). All results are dependent on the choice of the coordinate system: normalized poloidal flux {Psi}{sub N} and real space r/a. The most significant result, which was obtained with both parameter sets, shows a different scaling of the pedestal width for the electron temperature and the electron density. The presented scalings predict that in ITER and DEMO the temperature pedestal will be appreciably wider than the density pedestal. The pedestal top scaling for the pressure reveals differences between the electron and the ion pressure. In extrapolations this results in values for T{sub e,ped} of 4 keV (ITER) and 10 keV (DEMO), but significantly lower values for the ion temperature. A two-term method was applied to use the pedestal pressure to determine the pedestal contribution to the global confinement time {tau}{sub E}. The dependencies in the

  19. Electron thermal transport analysis in Tokamak à Configuration Variable

    Science.gov (United States)

    Asp, E.; Kim, J.-H.; Horton, W.; Porte, L.; Alberti, S.; Karpushov, A.; Martin, Y.; Sauter, O.; Turri, G.; Tcv Team

    2008-08-01

    A Tokamak à Configuration Variable (TCV) [G. Tonetti, A. Heym, F. Hofmann et al., in Proceedings of the 16th Symposium on Fusion Technology, London, U.K., edited by R. Hemsworth (North-Holland, Amsterdam, 1991), p. 587] plasma with high power density (up to 8MW/m3) core deposited electron cyclotron resonance heating at significant plasma densities (⩽7×1019m-3) is analyzed for the electron thermal transport. The discharge distinguishes itself as it has four distinct high confinement mode (H-mode) phases. An Ohmic H-mode with type III edge localized modes (ELMs), which turns into a type I ELMy H-mode when the ECRH is switched on. The ELMs then vanish, which gives rise to a quasistationary ELM-free H-mode. This ELM-free phase can be divided into two, one without magnetohydrodynamics (MHD) and one with. The MHD mode in the latter case causes the confinement to drop by ˜15%. For all four phases both large-scale trapped electron (TEM) and ion temperature gradient (ITG) modes and small-scale electron temperature gradient (ETG) modes are analyzed. The analytical TEM formulas have difficulty in explaining both the magnitude and the radial profile of the electron thermal flux. Collisionality governs the drive of the TEM, which for the discharge in question implies it can be driven by either the temperature or density gradient. The TEM response function is derived and it is shown to be relatively small and to have sharp resonances in its energy dependence. The ETG turbulence, predicted by the Institute for Fusion Studies electron gyrofluid code, is on the other hand driven solely by the electron temperature gradient. Both trapped and passing electrons add to the ETG instability and turbulent thermal flux. For easy comparison of the results of the above approaches and also with the Weiland model, a dimensionless error measure, the so-called average relative variance is introduced. According to this method the ETG model explains 70% of the variation in the electron heat

  20. Electromagnetic microinstabilities in tokamak plasmas using a global spectral approach

    Energy Technology Data Exchange (ETDEWEB)

    Falchetto, G. L

    2002-03-01

    Electromagnetic microinstabilities in tokamak plasmas are studied by means of a linear global eigenvalue numerical code. The code is the electromagnetic extension of an existing electrostatic global gyrokinetic spectral toroidal code, called GLOGYSTO. Ion dynamics is described by the gyrokinetic equation, so that ion finite Larmor radius effects are taken into account to all orders. Non adiabatic electrons are included in the model, with passing particles described by the drift-kinetic equation and trapped particles through the bounce averaged drift-kinetic equation. A low frequency electromagnetic perturbation is applied to a low -but finite- {beta}plasma (where the parameter {beta} identifies the ratio of plasma pressure to magnetic pressure); thus, the parallel perturbations of the magnetic field are neglected. The system is closed by the quasi-neutrality equation and the parallel component of Ampere's law. The formulation is applied to a large aspect ratio toroidal configuration, with circular shifted surfaces. Such a simple configuration enables one to derive analytically the gyrocenter trajectories. The system is solved in Fourier space, taking advantage of a decomposition adapted to the toroidal geometry. The major contributions of this thesis are as follows. The electromagnetic effects on toroidal Ion Temperature Gradient driven (ITG) modes are studied. The stabilization of these modes with increasing {beta}, as predicted in previous work, is confirmed. The inclusion of trapped electron dynamics enables the study of its coupling to the ITG modes and of Trapped Electron Modes (TEM) .The effects of finite {beta} are considered together with those of different magnetic shear profiles and of the Shafranov shift. The threshold for the destabilization of an electromagnetic mode is identified. Moreover, the global formulation yields for the first time the radial structure of this so-called Alfvenic Ion Temperature Gradient (AITG) mode. The stability of the

  1. Nonlinear effects of energetic particle driven instabilities in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Bruedgam, Michael

    2010-03-25

    In a tokamak plasma, a population of superthermal particles generated by heating methods can lead to a destabilization of various MHD modes. Due to nonlinear wave-particle interactions, a consequential fast particle redistribution reduces the plasma heating and can cause severe damages to the wall of the fusion device. In order to describe the wave-particle interaction, the drift-kinetic perturbative HAGIS code is applied which evolves the particle trajectories and the waves nonlinearly. For a simulation speed-up, the 6-d particle phase-space is reduced by the guiding centre approach to a 5-d description. The eigenfunction of the wave is assumed to be invariant, but its amplitude and phase is altered in time. A sophisticated {delta}/f-method is employed to model the change in the fast particle distribution so that numerical noise and the excessive number of simulated Monte-Carlo points are reduced significantly. The original code can only calculate the particle redistribution inside the plasma region. Therefore, a code extension has been developed during this thesis which enlarges the simulation region up to the vessel wall. By means of numerical simulations, this thesis addresses the problem of nonlinear waveparticle interactions in the presence of multiple MHD modes with significantly different eigenfrequencies and the corresponding fast particle transport inside the plasma. In this context, a new coupling mechanism between resonant particles and waves has been identified that leads to enhanced mode amplitudes and fast particle losses. The extension of the code provides for the first time the possibility of a quantitative and qualitative comparison between simulation results and recent measurements in the experiment. The findings of the comparison serve as a validation of both the theoretical model and the interpretation of the experimental results. Thus, a powerful interface tool has been developed for a deeper insight of nonlinear wave-particle interaction

  2. Improved Multi-Mode anomalous transport module for tokamak plasmas

    Science.gov (United States)

    Luo, L.; Rafiq, T.; Kritz, A. H.

    2013-10-01

    The Multi-Mode anomalous transport module version 7.1 (MMM7.1) is a theory-based transport model that is used to predict temperature, density and rotation profiles for tokamak plasmas in integrated whole device modeling codes. The theoretical foundation of the current version, MMM7.1, has been significantly advanced since the first released version in 1995, MMM95. The latest version of the Multi-Mode model, MMM7.1, includes an improved Weiland model for the ITG, TEM, and MHD modes, the Horton model for short wavelength ETG modes and the Rafiq model for the drift resistive inertial ballooning modes (DRIBMs). The ETG transport threshold in the Horton model is refined by using the threshold obtained from toroidal gyrokinetic ETG turbulence simulations. The different components of the MMM7.1 model provide contributions to transport in the different regions of plasma discharge. To facilitate the implementation of the latest version of the Multi-Mode module in integrated predictive modeling codes, a clearly specified interface is described and a test program is provided in order to examine the predictions provided by MMM7.1. MMM7.1 is documented and organized as a standalone module, which fully complies with the National Transport Code Collaboration (NTCC) standards. The MMM7.1 module has been used both with a standalone driver program as well as within the PTRANSP code. Results are presented to illustrate the extent to which the various component models contribute to transport in both L-mode and H-mode discharges.

  3. Nonlinear burn condition control in tokamaks using isotopic fuel tailoring

    Science.gov (United States)

    Boyer, Mark D.; Schuster, Eugenio

    2015-08-01

    One of the fundamental problems in tokamak fusion reactors is how to control the plasma density and temperature in order to regulate the amount of fusion power produced by the device. Control of these parameters will be critical to the success of burning plasma experiments like ITER. The most previous burn condition control efforts use either non-model based control designs or techniques based on models linearized around particular operating points. Such strategies limit the potential operational space and must be carefully retuned or redesigned to accommodate changes in operating points or plasma parameters. In this work, a nonlinear dynamic model of the spatial averages of energy and ion species densities is used to synthesize a nonlinear feedback controller for stabilizing the burn condition. The nonlinear model-based control strategy guarantees a much larger operational space than previous linear controllers. Because it is not designed around a particular operating point, the controller can be used to move from one burn condition to another. The proposed scheme first attempts to use regulation of the auxiliary heating power to reject temperature perturbations, then, if necessary, uses isotopic fuel tailoring as a way to reduce fusion heating during positive temperature perturbations. A global model of hydrogen recycling is incorporated into the model used for design and simulation, and the proposed control scheme is tested for a range of recycling model parameters. As we find the possibility of changing the isotopic mix can be limited for certain unfavorable recycling conditions, we also consider impurity injection as a back-up method for controlling the system. A simple supervisory control strategy is proposed to switch between the primary and back-up control schemes based on stability and performance criteria. A zero-dimensional simulation study is used to study the performance of the control scheme for several scenarios and model parameters. Finally, a one

  4. Impurity effects on trapped electron mode in tokamak plasmas

    Science.gov (United States)

    Du, Huarong; Wang, Zheng-Xiong; Dong, J. Q.

    2016-07-01

    The effects of impurity ions on the trapped electron mode (TEM) in tokamak plasmas are numerically investigated with the gyrokinetic integral eigenmode equation. It is shown that in the case of large electron temperature gradient ( η e ), the impurity ions have stabilizing effects on the TEM, regardless of peaking directions of their density profiles for all normalized electron density gradient R / L n e . Here, R is the major radius and L n e is the electron density gradient scale length. In the case of intermediate and/or small η e , the light impurity ions with conventional inwardly (outwardly) peaked density profiles have stabilizing effects on the TEM for large (small) R / L n e , while the light impurity ions with steep inwardly (outwardly) peaked density profiles can destabilize the TEM for small (large) R / L n e . Besides, the TEM driven by density gradient is stabilized (destabilized) by the light carbon or oxygen ions with inwardly (outwardly) peaked density profiles. In particular, for flat and/or moderate R / L n e , two independent unstable modes, corresponding respectively to the TEM and impurity mode, are found to coexist in plasmas with impurity ions of outwardly peaked density profiles. The high Z tungsten impurity ions play a stronger stabilizing role in the TEM than the low Z impurity ions (such as carbon and oxygen) do. In addition, the effects of magnetic shear and collision on the TEM instability are analyzed. It is shown that the collisionality considered in this work weakens the trapped electron response, leading to a more stable TEM instability, and that the stabilizing effects of the negative magnetic shear on the TEM are more significant when the impurity ions with outwardly peaked density profile are taken into account.

  5. Spectroscopic system for impurity measurements in the TJ-1 Tokamak of JEN; Un sistema espectroscopico para medidas de impurezas en el Tokamak TJ-1 de la JEN

    Energy Technology Data Exchange (ETDEWEB)

    Navas, G.; Zurro, B.

    1982-07-01

    we describe a spectroscopic system with spatial resolution capability that has been configured for plasma diagnostic in the TJ-1 Tokamak of JEN. The experimental system, based on a one meter monochromator, has been absolutely calibrated using a tungsten-halogen lamp. The calibration procedures and the absolute spectral sensitivity are presented as well as its dependence with the polarization. A simplified spectroscopic model of the radiation emitted by the intrinsic plasma impurities (C, 0, . . . ) has been developed. A one dimensional model of the temporal evolution of various ionization stages in coronal equilibrium is used to predict the electron temperature and impurity concentration. This model has been applied to experimental data from several Tokamaks. (Author) 23 refs.

  6. The Effect of Equilibrium Current Profiles on MHD Instabilities in Tokamaks%The Effect of Equilibrium Current Profiles on MHD Instabilities in Tokamaks

    Institute of Scientific and Technical Information of China (English)

    李莉; 刘悦; 许欣洋; 夏新念

    2012-01-01

    A cylindrical model of linear MHD instabilities in tokamaks is presented. In the model, the cylindrical plasma is surrounded by a vacuum which is divided into inner and outer vacuum areas by a conducting wall. Linearized resistivity MHD equations with plasma viscosity are adopted to describe our model, and the equations are solved numerically as an initial value problem. Some of the results are used as benchmark tests for the code, and then a series of equilibrium current profiles are used to simulate the bootstrap current profiles in actual experiments with a bump on tail. Thus the effects of these kinds of profiles on MHD instabilities in tokamaks are revealed. From the analysis of the numerical results, it is found that more plasma can be confined when the center of the current bump is closer to the plasma surface, and a higher and narrower current bump has a better stabilizing effect on the MHD instabilities.

  7. Study of the electron heat transport in Tore-Supra tokamak; Etude du transport de la chaleur electronique dans le Tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Harauchamps, E

    2004-07-01

    This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)

  8. Method of compensation spires for the detection of the diamagnetic effect in a Tokamak; Metodo de espiras de compensacion para la deteccion del efecto diamagnetico en un Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Colunga S, S

    1990-09-15

    In this report the classical detection method of the diamagnetic effect by means of a rolled spire on the discharges chamber in the poloidal direction and the difficulties related with this are analyzed. An alternative method that increases considerably the detection sensibility of the diamagnetic effect and that for its simplicity it is quite attractive for its application to the Tokamak Novillo of the ININ is presented. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

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

  10. Turbulence in tokamak plasmas. Effect of a radial electric field shear; Turbulence dans les plasmas de tokamaks. Effet d`un cisaillement de champ electrique radial

    Energy Technology Data Exchange (ETDEWEB)

    Payan, J.

    1994-05-01

    After a review of turbulence and transport phenomena in tokamak plasmas and the radial electric field shear effect in various tokamaks, experimental measurements obtained at Tore Supra by the means of the ALTAIR plasma diagnostic technique, are presented. Electronic drift waves destabilization mechanisms, which are the main features that could describe the experimentally observed microturbulence, are then examined. The effect of a radial electric field shear on electronic drift waves is then introduced, and results with ohmic heating are studied together with relations between turbulence and transport. The possible existence of ionic waves is rejected, and a spectral frequency modelization is presented, based on the existence of an electric field sheared radial profile. The position of the inversion point of this field is calculated for different values of the mean density and the plasma current, and the modelization is applied to the TEXT tokamak. The radial electric field at Tore Supra is then estimated. The effect of the ergodic divertor on turbulence and abnormal transport is then described and the density fluctuation radial profile in presence of the ergodic divertor is modelled. 80 figs., 120 refs.

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

    Science.gov (United States)

    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.

  12. Identification of the ubiquitous Coriolis momentum pinch in JET tokamak plasmas

    NARCIS (Netherlands)

    Weisen, H.; Camenen, Y.; Salmi, A.; Versloot, T. W.; de Vries, P. C.; Maslov, M.; Tala, T.; Beurskens, M.; Giroud, C.

    2012-01-01

    A broad survey of the experimental database of neutral beam heated plasmas in the JET tokamak has established the theoretically expected ubiquity, in rotating plasmas, of a convective transport mechanism which has its origin in the vertical particle drift resulting from the Coriolis force. This inwa

  13. Localization of the magnetic reconnection zone during sawtooth crashes in tokamak plasmas

    NARCIS (Netherlands)

    Munsat, T.; Park, H. K.; Classen, I.G.J.; Domier, C.W.; Donne, A. J. H.; N C Luhmann Jr.,; Mazzucato, E.; van de Pol, M.J.

    2007-01-01

    Recent 2D spatially and temporally resolved measurements of electron temperature fluctuations in the tokamak core have revealed new information on the dynamics of the sawtooth crash. Measures of poloidal localization of the reconnection zone are achieved through direct analysis of the 2D data and th

  14. Real-time optical plasma boundary reconstruction for plasma position control at the TCV Tokamak

    NARCIS (Netherlands)

    Hommen, G.; Baar, M. de; Duval, B.P.; Andrebe, Y.; Le, H.B.; Klop, M.A.; Doelman, N.J.; Witvoet, G.; Steinbuch, M.

    2014-01-01

    A dual, high speed, real-time visible light camera setup was installed on the TCV tokamak to reconstruct optically and in real-time the plasma boundary shape. Localized light emission from the plasma boundary in tangential view, broadband visible images results in clearly resolved boundary edge-feat

  15. Fast-ion dynamics in the TEXTOR tokamak measured by collective Thomson scattering

    DEFF Research Database (Denmark)

    Bindslev, H.; Nielsen, S.K.; Porte, L.;

    2006-01-01

    Here we present the first measurements by collective Thomson scattering of the evolution of fast-ion populations in a magnetically confined fusion plasma. 150 kW and 110 Ghz radiation from a gyrotron were scattered in the TEXTOR tokamak plasma with energetic ions generated by neutral beam injecti...

  16. Fusion Energy-Production from a Deuterium-Tritium Plasma in the Jet Tokamak

    NARCIS (Netherlands)

    Rebut, P. H.; Gibson, A.; Huguet, M.; Adams, J. M.; Alper, B.; Altmann, H.; Andersen, A.; Andrew, P.; Angelone, M.; Aliarshad, S.; Baigger, P.; Bailey, W.; Balet, B.; Barabaschi, P.; Barker, P.; Barnsley, R.; Baronian, M.; Bartlett, D. V.; Baylor, L.; Bell, A. C.; Benali, G.; Bertoldi, P.; Bertolini, E.; Bhatnagar, V.; Bickley, A. J.; Binder, D.; Bindslev, H.; Bonicelli, T.; Booth, S. J.; Bosia, G.; Botman, M.; Boucher, D.; Boucquey, P.; Breger, P.; Brelen, H.; Brinkschulte, H.; Brooks, D.; Brown, A.; Brown, T.; Brusati, M.; Bryan, S.; Brzozowski, J.; Buchse, R.; Budd, T.; Bures, M.; Businaro, T.; Butcher, P.; Buttgereit, H.; Caldwellnichols, C.; Campbell, D. J.; Card, P.; Celentano, G.; Challis, C. D.; Chankin, A. V.; Cherubini, A.; Chiron, D.; Christiansen, J.; Chuilon, P.; Claesen, R.; Clement, S.; Clipsham, E.; Coad, J. P.; Coffey, I. H.; Colton, A.; Comiskey, M.; Conroy, S.; Cooke, M.; Cooper, D.; Cooper, S.; Cordey, J. G.; Core, W.; Corrigan, G.; Corti, S.; Costley, A. E.; Cottrell, G.; Cox, M.; Cripwell, P.; Dacosta, O.; Davies, J.; Davies, N.; de Blank, H.; De Esch, H.; Dekock, L.; Deksnis, E.; Delvart, F.; Dennehinnov, G. B.; Deschamps, G.; Dickson, W. J.; Dietz, K. J.; Dmitrenko, S. L.; Dmitrieva, M.; Dobbing, J.; Doglio, A.; Dolgetta, N.; Dorling, S. E.; Doyle, P. G.; Duchs, D. F.; Duquenoy, H.; Edwards, A.; Ehrenberg, J.; Ekedahl, A.; Elevant, T.; Erents, S.K.; Eriksson, L. G.; Fajemirokun, H.; Falter, H.; Freiling, J.; Freville, F.; Froger, C.; Froissard, P.; Fullard, K.; Gadeberg, M.; Galetsas, A.; Gallagher, T.; Gambier, D.; Garribba, M.; Gaze, P.; Giannella, R.; Gill, R. D.; Girard, A.; Gondhalekar, A.; Goodall, D.; Gormezano, C.; Gottardi, N. A.; Gowers, C.; Green, B. J.; Grievson, B.; Haange, R.; Haigh, A.; Hancock, C. J.; Harbour, P. J.; Hartrampf, T.; Hawkes, N. C.; Haynes, P.; Hemmerich, J. L.; Hender, T.; Hoekzema, J.; Holland, D.; Hone, M.; Horton, L.; How, J.; Huart, M.; Hughes, I.; Hughes, T. P.; Hugon, M.; Huo, Y.; Ida, K.; Ingram, B.; Irving, M.; Jacquinot, J.; Jaeckel, H.; Jaeger, J. F.; Janeschitz, G.; Jankovicz, Z.; Jarvis, O. N.; Jensen, F.; Jones, E. M.; Jones, H. D.; Jones, Lpdf; Jones, S.; Jones, T. T. C.; Junger, J. F.; Junique, F.; Kaye, A.; Keen, B. E.; Keilhacker, M.; Kelly, G. J.; Kerner, W.; Khudoleev, A.; Konig, R.; Konstantellos, A.; Kovanen, M.; Kramer, G.; Kupschus, P.; Lasser, R.; Last, J. R.; Laundy, B.; Laurotaroni, L.; Laveyry, M.; Lawson, K.; Lennholm, M.; Lingertat, J.; Litunovski, R. N.; Loarte, A.; Lobel, R.; Lomas, P.; Loughlin, M.; Lowry, C.; Lupo, J.; Maas, A. C.; Machuzak, J.; Macklin, B.; Maddison, G.; Maggi, C. F.; Magyar, G.; Mandl, W.; Marchese, V.; Marcon, G.; Marcus, F.; Mart, J.; Martin, D.; Martin, E.; Martinsolis, R.; Massmann, P.; Matthews, G.; McBryan, H.; McCracken, G.; McKivitt, J.; Meriguet, P.; Miele, P.; Miller, A.; Mills, J.; Mills, S. F.; Millward, P.; Milverton, P.; Minardi, E.; Mohanti, R.; Mondino, P. L.; Montgomery, D.; Montvai, A.; Morgan, P.; Morsi, H.; Muir, D.; Murphy, G.; Myrnas, R.; Nave, F.; Newbert, G.; Newman, M.; Nielsen, P.; Noll, P.; Obert, W.; Obrien, D.; Orchard, J.; Orourke, J.; Ostrom, R.; Ottaviani, M.; Pain, M.; Paoletti, F.; Papastergiou, S.; Parsons, W.; Pasini, D.; Patel, D.; Peacock, A.; Peacock, N.; Pearce, R. J. M.; Pearson, D.; Peng, J. F.; Desilva, R. P.; Perinic, G.; Perry, C.; Petrov, M.; Pick, M. A.; Plancoulaine, J.; Poffe, J. P.; Pohlchen, R.; Porcelli, F.; Porte, L.; Prentice, R.; Puppin, S.; Putvinskii, S.; Radford, G.; Raimondi, T.; Deandrade, M. C. R.; Reichle, R.; Reid, J.; Richards, S.; Righi, E.; Rimini, F.; Robinson, D.; Rolfe, A.; Ross, R. T.; Rossi, L.; Russ, R.; Rutter, P.; Sack, H. C.; Sadler, G.; Saibene, G.; Salanave, J. L.; Sanazzaro, G.; Santagiustina, A.; Sartori, R.; Sborchia, C.; Schild, P.; Schmid, M.; Schmidt, G.; Schunke, B.; Scott, S. M.; Serio, L.; Sibley, A.; Simonini, R.; Sips, A.C.C.; Smeulders, P.; Smith, R.; Stagg, R.; Stamp, M.; Stangeby, P.; Stankiewicz, R.; Start, D. F.; Steed, C. A.; Stork, D.; Stott, P.E.; Stubberfield, P.; Summers, D.; Summers, H.; Svensson, L.; Tagle, J. A.; Talbot, M.; Tanga, A.; Taroni, A.; Terella, C.; Terrington, A.; Tesini, A.; Thomas, P. R.; Thompson, E.; Thomsen, K.; Tibone, F.; Tiscornia, A.; Trevalion, P.; Tubbing, B.; Vanbelle, P.; Vanderbeken, H.; Vlases, G.; von Hellermann, M.; Wade, T.; Walker, C.; Walton, R.; Ward, D.; Watkins, M. L.; Watkins, N.; Watson, M. J.; Weber, S.; Wesson, J.; Wijnands, T. J.; Wilks, J.; Wilson, D.; Winkel, T.; Wolf, R.; Wong, D.; Woodward, C.; Wu, Y.; Wykes, M.; Young, D.; Young, I. D.; Zannelli, L.; Zolfaghari, A.; Zwingmann, W.

    1992-01-01

    The paper describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixtures. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to

  17. Proto-CIRCUS Tilted-Coil Tokamak-Torsatron Hybrid: Design and Construction

    CERN Document Server

    Clark, A W; Hammond, K C; Kornbluth, Y; Spong, D A; Sweeney, R; Volpe, F A

    2014-01-01

    We present the field-line modeling, design and construction of a prototype circular-coil tokamak-torsatron hybrid called Proto-CIRCUS. The device has a major radius R = 16 cm and minor radius a < 5 cm. The six "toroidal field" coils are planar as in a tokamak, but they are tilted. This, combined with induced or driven plasma current, is expected to generate rotational transform, as seen in field-line tracing and equilibrium calculations. The device is expected to operate at lower plasma current than a tokamak of comparable size and magnetic field, which might have interesting implications for disruptions and steady-state operation. Additionally, the toroidal magnetic ripple is less pronounced than in an equivalent tokamak in which the coils are not tilted. The tilted coils are interlocked, resulting in a relatively low aspect ratio, and can be moved, both radially and in tilt angle, between discharges. This capability will be exploited for detailed comparisons between calculations and field-line mapping me...

  18. A comparison of steady-state ARIES and pulsed PULSAR tokamak power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.

    1994-07-01

    The multi-institutional ARIES study has completed a series of three steady-state and two pulsed cost-optimized conceptual designs of commercial tokamak fusion power plants that vary the level of assumed advances in technology and physics. The cost benefits of various design options are compared quantitatively. Possible means to improve the economic competitiveness of fusion are suggested.

  19. High spatial and temporal resolution charge exchange recombination spectroscopy on the HL-2A tokamak

    NARCIS (Netherlands)

    Wei, Y. L.; Yu, D. L.; Liu, L.; Ida, K.; von Hellermann, M.; Cao, J. Y.; Sun, A. P.; Ma, Q.; Chen, W. J.; Liu, Yi; Yan, L. W.; Yang, Q. W.; Duan, X. R.; Liu, Yong

    2014-01-01

    A 32/64-channel charge exchange recombination spectroscopy (CXRS) diagnostic system is developed on the HL-2A tokamak (R = 1.65 m, a = 0.4 m), monitoring plasma ion temperature and toroidal rotation velocity simultaneously. A high throughput spectrometer (F/2.8) and a pitch-controlled fiber bundle e

  20. New dual gas puff imaging system with up-down symmetry on experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Shao, L. M.; Zweben, S. J.;

    2012-01-01

    advanced superconducting tokamak (EAST). The two views are up-down symmetric about the midplane and separated by a toroidal angle of 66.6 degrees. A linear manifold with 16 holes apart by 10 mm is used to form helium gas cloud at the 130x130 mm (radial versus poloidal) objective plane. A fast camera...

  1. Commissioning of electron cyclotron emission imaging instrument on the DIII-D tokamak and first data

    NARCIS (Netherlands)

    Tobias, B.; Domier, C.W.; Liang, T.; Kong, X.; Yu, L.; Yun, G. S.; Park, H. K.; Classen, I.G.J.; Boom, J. E.; Donne, A. J. H.; Munsat, T.; Nazikian, R.; Van Zeeland, M.; Boivin, R. L.; N C Luhmann Jr.,

    2010-01-01

    A new electron cyclotron emission imaging diagnostic has been commissioned on the DIII-D tokamak. Dual detector arrays provide simultaneous two-dimensional images of T-e fluctuations over radially distinct and reconfigurable regions, each with both vertical and radial zoom capability. A total of 320

  2. L-H transition in the mega-Amp spherical tokamak

    DEFF Research Database (Denmark)

    Akers, R.J.; Counsell, G.F.; Sykes, A.

    2002-01-01

    H-mode plasmas have been achieved on the MAST spherical tokamak at input power considerably higher than predicted by conventional threshold scalings. Following L-H transition, a clear improvement in energy confinement is obtained, exceeding recent international scalings even at densities approach...

  3. 3D-MAPTOR Code for computation of magnetic fields in tokamaks

    Science.gov (United States)

    ChÁVez-AlarcÓN, Esteban; Herrera-VelÁZquez, Julio

    2009-11-01

    A 3dimensional code has been developed in order to determine the magnetic field in tokamaks, starting from the assumption that the toroidal and vertical field coils are all circular, as well as the cross section of the plasma current distribution. It was earlier used to study the stochastization of the outer magnetic surfaces [1] and to reconstruct the evolution of the plasma column, using the experimental signals of tokamak discharges. These results were compared with tomographic reconsructions of the ISTTOK tokamak [2]. We present an upgrade of the code, in which rectangular toroidal field coils and D shaped plasma current cross sections can be included. The code is particularly useful to study the effect of the ripple along the toroidal coordinate.[4pt] [1] E. Ch'avez, et al., ``Stochastization of Magnetic Field Surfaces in Tokamaks by an Inner Coil'' in Plasma and Fusion Science, AIP Conference Proceedings Series 875 (2006) pp.347-349.[0pt] [2] B.B. Carvalho, et al., ``Real-time plasma control based on the ISTTOK tomography diagnostic'', Rev. Sci. Instrum. 79 (2008)10F329.

  4. EAST gets a head start in steady-state tokamak physics

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    @@ While the International Thermonuclear Experimental Reactor (ITER) project is underway at Cadarache, France, an initiative in building the next-generation tokamak at the CAS Institute of Plasma Physics (ASIPP) in Hefei, capital of east China's Anhui Province, offers crucial expertise.

  5. Development on systems configuration in ITER Tokamak Complex and Auxiliary Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kuehn, Ingo, E-mail: ingo.kuehn@iter.org [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Kotamaki, Miikka [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Schmieder, Laurent [Fusion for Energy (F4E), CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Cordier, Jean-Jacques; Chiocchio, Stefano; Carafa, Leontin; Klingsmith, James; Patisson, Laurent; Rigoni, Giuliano [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Tsedri, Thibault [AREVA CNIM KAH System Engineering Support, CS 50497, 13593 Aix en Provence Cedex 3 (France)

    2011-10-15

    The ITER site consists of almost 30 buildings to service the Tokamak machine which is located in the centre of the Tokamak Complex facility with the Tokamak-, Diagnostic- and Tritium building. The design of a large part of the ITER plant systems will be executed by the ITER Domestic Agencies or their industrial suppliers under functional specifications provided by the ITER Organization. At the same time, the detailed design of the building is carried out by the European Domestic Agency 'Fusion for Energy' (F4E). In order to allow an efficient identification of the ITER configuration as well as to manage the concurrent engineering activities and to simplify the identification and assessment of changes, the design of each ITER plant systems is described in the so-called Configuration Management Models (CMM). These are light CATIA 3D models that define the required space envelope and the physical interfaces in-between the systems and the buildings. The paper describes the procedure adopted for the control of the baseline configuration of the Tokamak Complex facility and Auxiliary Buildings with their associated plant systems and illustrates the current status as well as recent developments in the different systems.

  6. Advanced Tokamak current density profiles for non-inductive Tore Supra operation

    Energy Technology Data Exchange (ETDEWEB)

    Kazarian-Vibert, F.; Litaudon, X.; Arslanbekov, R.; Bibet, P.; Froissard, P.; Goniche, M.; Hoang, G.T.; Joffrin, E.; Moreau, D.; Peysson, Y.; Rey, G. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    1995-12-31

    This document deals with the advanced Tokamak concept concerning self consistent hollow current density profiles. Several Lower Hybrid experiments performed on Tore Supra are presented: the feasibility of the constant-flux operation mode is demonstrated and a new improved confinement regime with a reversed shear has been obtained. (TEC). 12 refs., 5 figs.

  7. Islands of Runaway Electrons in the Textor Tokamak and Relation to Transport in a Stochastic Field

    NARCIS (Netherlands)

    R. Jaspers,; Cardozo, N. J. L.; Finken, K.H.; Schokker, B. C.; Mank, G.; Fuchs, G.; Schüller, F. C.

    1994-01-01

    A population of 30 MeV runaway electrons in the TEXTOR tokamak is diagnosed by their synchrotron emission. During pellet injection a large fraction of the population is lost within 600 mus. This rapid loss is attributed to stochastization of the magnetic field. The remaining runaways form a narrow,

  8. Scale size of magnetic turbulence in tokamaks probed with 30-MeV electrons

    NARCIS (Netherlands)

    Entrop, I.; Cardozo, N. J. L.; R. Jaspers,; Finken, K.H.

    2000-01-01

    Measurements of synchrotron radiation emitted by 30-MeV runaway electrons in the TEXTOR-94 tokamak show that the runaway population decays after switching on neutral beam injection (NBI). The decay starts only with a significant delay, which decreases with increasing NBI heating power. This delay pr

  9. Integrated modeling of temperature profiles in L-mode tokamak discharges

    Science.gov (United States)

    Rafiq, T.; Kritz, A. H.; Tangri, V.; Pankin, A. Y.; Voitsekhovitch, I.; Budny, R. V.

    2014-12-01

    Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

  10. Integrated modeling of temperature profiles in L-mode tokamak discharges

    Energy Technology Data Exchange (ETDEWEB)

    Rafiq, T.; Kritz, A. H.; Tangri, V. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Voitsekhovitch, I. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Budny, R. V. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2014-12-15

    Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

  11. Localized bulk electron heating with ICRF mode conversion in the JET tokamak

    DEFF Research Database (Denmark)

    Mantsinen, M.J.; Mayoral, M.-L.; Eester, D. Van

    2004-01-01

    Ion cyclotron resonance frequencies (ICRF) mode conversion has been developed for localized on-axis and off-axis bulk electron heating on the JET tokamak. The fast magnetosonic waves launched from the low-field side ICRF antennas are mode-converted to short-wavelength waves on the high-field side...

  12. Review Committee report on the conceptual design of the Tokamak Physics Experiment

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    This report discusses the following topics on the conceptual design of the Tokamak Physics Experiment: Role and mission of TPX; overview of design; physics design assessment; engineering design assessment; evaluation of cost, schedule, and management plans; and, environment safety and health.

  13. Measurements of Intrinsic Ion Bernstein Waves in a Tokamak by Collective Thomson Scattering

    DEFF Research Database (Denmark)

    Korsholm, Søren Bang; Stejner Pedersen, Morten; Bindslev, Henrik;

    2011-01-01

    In this Letter we report measurements of collective Thomson scattering (CTS) spectra with clear signatures of ion Bernstein waves and ion cyclotron motion in tokamak plasmas. The measured spectra are in accordance with theoretical predictions and show clear sensitivity to variation in the density...

  14. A Study on Main Breaker for Quench Protection of HT-7U Toroidal Superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

    许留伟; 刘小宁

    2002-01-01

    This paper proposes a quench protection project of HT-7U toroidal superconducting tokamak through a forced commutation analysis of DC circuit breaker (DCCB) paralleling fuse.Based on the requirement of quench protection, main parameters are selected. Experimental results demonstrate the validity of this proposed project.

  15. Nonlocal transient transport and thermal barriers in Rijnhuizen Tokamak project plasmas

    NARCIS (Netherlands)

    Mantica, P.; Galli, P.; Gorini, G.; Hogeweij, G. M. D.; de Kloe, J.; Cardozo, N. J. L.; R. T. P. Team,

    1999-01-01

    In the Rijnhuizen Tokamak Project plasmas, a transient rise of the core electron temperature is observed when hydrogen pellets are injected tangentially to induce fast cooling of the peripheral region. High-resolution Thomson scattering measurements show that the T-e rise is associated with large te

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

    Science.gov (United States)

    Leddy, J.; Dudson, B.; Romanelli, M.; Shanahan, B.; Walkden, N.

    2017-03-01

    Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are ;closed; (i.e. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry can be matched in the poloidal plane while maintaining a field-aligned coordinate. This system is implemented in BOUT++ and tested for accuracy using the method of manufactured solutions. A MAST edge cross-section is simulated using a fluid plasma model and the results show expected behaviour for density, temperature, and velocity. Finally, simulations of an isolated divertor leg are conducted with and without neutrals to demonstrate the ion-neutral interaction near the divertor plate and the corresponding beneficial decrease in plasma temperature.

  17. Structures in T-e profiles: High resolution Thomson scattering in the Rijnhuizen tokamak project

    NARCIS (Netherlands)

    Beurskens, M. N. A.; Barth, C. J.; Cardozo, N. J. L.; van der Meiden, H. J.; R. T. P. Team,

    1999-01-01

    In the Rijnhuizen tokamak project, the double pulse multiposition Thomson scattering diagnostic is in full operation. Its high spatial resolution enables the measurement of small scale structures in T-e, n(e), and p(e). Thomson scattering profiles during an ordinary sawtooth crash show the displacem

  18. High resolution equilibrium calculations of pedestal and SOL plasma in tokamaks

    Science.gov (United States)

    Medvedev, S. Yu; Martynov, A. A.; Drozdov, V. V.; Ivanov, A. A.; Poshekhonov, Yu Yu

    2017-02-01

    For integrated modeling of equilibrium, stability and dynamics of the divertor tokamak plasma with scrape-off layer (SOL) high resolution equilibrium calculations are needed. A new version of the CAXE equilibrium code computes the tokamak equilibrium on a numerical grid adaptive to magnetic surfaces both in the plasma region with closed flux surfaces and in the SOL region with open magnetic lines. The plasma profiles can be prescribed independently in each region with nested flux surfaces, and realistic SOL profiles with very short pressure drop off length can be accurately treated. The influence of the finite current density in SOL on the connection length is studied. From the point of view of the MHD equilibrium and stability modeling, self-consistent calculations of diverted tokamak configurations with finite current density at the separatrix require taking into account plasma outside the separatrix. Calculated high resolution equilibria provide an input to new versions of the ideal MHD stability codes treating tokamak plasma with SOL. The study of the influence of the pressure gradient profile in the pedestal plasma inside and outside the separatrix on the pedestal height limit set by external kink-ballooning mode stability is presented. Another possible application of the high resolution pedestal and SOL equilibrium code is a coupling to the SOLPS code with a purpose to increase equilibrium accuracy and support self-consistent plasma flow/equilibrium modeling.

  19. Langmuir-magnetic probe measurements of ELMs and dithering cycles in the EAST tokamak

    DEFF Research Database (Denmark)

    Yan, Ning; Naulin, Volker; Xu, G. S.

    2014-01-01

    Measurements of the dynamical behavior associated with edge localized modes (ELMs) have been carried out in the Experimental Advanced Superconducting Tokamak (EAST) by direct probing near the separatrix and far scrape-off layer (SOL) using electrostatic as well as magnetic probes. Type-III ELMs...

  20. Existence Conditions of Collision Sheath in the Tokamak Scrape-off Layer

    Institute of Scientific and Technical Information of China (English)

    GAOQingdi; CHENXiaoping

    2002-01-01

    In controlled nuclear fusion devices like tokamaks, plasma particles are confined by closed magnetic flux surfaces. Outside the last closed flux surface (LCFS), plasma is in direct contact with a solid wall in the scrape-off layer (SOL). Due to the different

  1. Development of real-time plasma analysis and control algorithms for the TCV tokamak using Simulink

    NARCIS (Netherlands)

    Felici, F.; Le, H. B.; J. I. Paley,; Duval, B. P.; Coda, S.; Moret, J. M.; Bortolon, A.; L. Federspiel,; Goodman, T. P.; Hommen, G.; A. Karpushov,; Piras, F.; A. Pitzschke,; J. Romero,; G. Sevillano,; Sauter, O.; Vijvers, W.; TCV team,

    2014-01-01

    One of the key features of the new digital plasma control system installed on the TCV tokamak is the possibility to rapidly design, test and deploy real-time algorithms. With this flexibility the new control system has been used for a large number of new experiments which exploit TCV's powerful

  2. Fast-ion transport induced by Alfvén eigenmodes in the ASDEX Upgrade tokamak

    DEFF Research Database (Denmark)

    Garcia-Munoz, M.; Classen, I.G.J.; Geiger, B.

    2011-01-01

    A comprehensive suite of diagnostics has allowed detailed measurements of the Alfvén eigenmode (AE) spatial structure and subsequent fast-ion transport in the ASDEX Upgrade (AUG) tokamak [1]. Reversed shear Alfvén eigenmodes (RSAEs) and toroidal induced Alfvén eigenmodes (TAEs) have been driven u...

  3. Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

    DEFF Research Database (Denmark)

    Meyer, H.; Eich, T.; Beurskens, M.

    2017-01-01

    Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine ...

  4. The European Integrated Tokamak Modelling (ITM) effort: achievements and first physics results

    NARCIS (Netherlands)

    G.L. Falchetto,; Coster, D.; Coelho, R.; Scott, B. D.; Figini, L.; Kalupin, D.; Nardon, E.; Nowak, S.; L.L. Alves,; Artaud, J. F.; Basiuk, V.; João P.S. Bizarro,; C. Boulbe,; Dinklage, A.; Farina, D.; B. Faugeras,; Ferreira, J.; Figueiredo, A.; Huynh, P.; Imbeaux, F.; Ivanova-Stanik, I.; Jonsson, T.; H.-J. Klingshirn,; Konz, C.; Kus, A.; Marushchenko, N. B.; Pereverzev, G.; M. Owsiak,; Poli, E.; Peysson, Y.; R. Reimer,; Signoret, J.; Sauter, O.; Stankiewicz, R.; Strand, P.; Voitsekhovitch, I.; Westerhof, E.; T. Zok,; Zwingmann, W.; ITM-TF contributors,; ASDEX Upgrade team,; JET-EFDA Contributors,

    2014-01-01

    A selection of achievements and first physics results are presented of the European Integrated Tokamak Modelling Task Force (EFDA ITM-TF) simulation framework, which aims to provide a standardized platform and an integrated modelling suite of validated numerical codes for the simulation and

  5. Energetic-ion-driven global instabilities in stellarator/helical plasmas and comparison with tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Toi, K. [National Institute for Fusion Science, Toki, Japan; Ogawa, K. [Nagoya University, Japan; Isobe, M. [National Institute for Fusion Science, Toki, Japan; Osakabe, M. [National Institute for Fusion Science, Toki, Japan; Spong, Donald A [ORNL; Todo, Yasushi [National Institute for Fusion Science, Toki, Japan

    2011-01-01

    Comprehensive understanding of energetic-ion-driven global instabilities such as Alfven eigenmodes (AEs) and their impact on energetic ions and bulk plasma is crucially important for tokamak and stellarator/helical plasmas and in the future for deuterium-tritium (DT) burning plasma experiments. Various types of global modes and their associated enhanced energetic ion transport are commonly observed in toroidal plasmas. Toroidicity-induced AEs and ellipticity-induced AEs, whose gaps are generated through poloidal mode coupling, are observed in both tokamak and stellarator/helical plasmas. Global AEs and reversed shear AEs, where toroidal couplings are not as dominant were also observed in those plasmas. Helicity induced AEs that exist only in 3D plasmas are observed in the large helical device (LHD) and Wendelstein 7 Advanced Stellarator plasmas. In addition, the geodesic acoustic mode that comes from plasma compressibility is destabilized by energetic ions in both tokamak and LHD plasmas. Nonlinear interaction of these modes and their influence on the confinement of the bulk plasma as well as energetic ions are observed in both plasmas. In this paper, the similarities and differences in these instabilities and their consequences for tokamak and stellarator/helical plasmas are summarized through comparison with the data sets obtained in LHD. In particular, this paper focuses on the differences caused by the rotational transform profile and the 2D or 3D geometrical structure of the plasma equilibrium. Important issues left for future study are listed.

  6. Consideration of neutral beam prompt loss in the design of a tokamak helicon antenna

    Energy Technology Data Exchange (ETDEWEB)

    Pace, D.C., E-mail: pacedc@fusion.gat.com; Van Zeeland, M.A.; Fishler, B.; Murphy, C.

    2016-11-15

    Highlights: • Neutral beam prompt losses place appreciable power on an in-vessel tokamak antenna. • Simulations predict prompt loss power and inform protective tile design. • Experiments confirm the validity of the prompt loss simulations. - Abstract: Neutral beam prompt losses (injected neutrals that ionize such that their first poloidal transit intersects with the wall) can put appreciable power on the outer wall of tokamaks, and this power may damage the wall or other internal components. These prompt losses are simulated including a protruding helicon antenna installation in the DIII-D tokamak and it is determined that 160 kW of power will impact the antenna during the injection of a particular neutral beam. Protective graphite tiles are designed in response to this modeling and the wall shape of the installed antenna is precisely measured to improve the accuracy of these calculations. Initial experiments confirm that the antenna component temperature increases according to the amount of neutral beam energy injected into the plasma. In this case, only injection of beams that are aimed counter to the plasma current produce an appreciable power load on the outer wall, suggesting that the effect is of little concern for tokamaks featuring only co-current neutral beam injection. Incorporating neutral beam prompt loss considerations into the design of this in-vessel component serves to ensure that adequate protection or cooling is provided.

  7. Modeling of Resistive Wall Modes in Tokamak and Reversed Field Pinch Configurations of KTX

    Science.gov (United States)

    Han, Rui; Zhu, Ping; Bai, Wei; Lan, Tao; Liu, Wandong

    2016-10-01

    Resistive wall mode is believed to be one of the leading causes for macroscopic degradation of plasma confinement in tokamaks and reversed field pinches (RFP). In this study, we evaluate the linear RWM instability of Keda Torus eXperiment (KTX) in both tokamak and RFP configurations. For the tokamak configuration, the extended MHD code NIMROD is employed for calculating the dependence of the RWM growth rate on the position and conductivity of the vacuum wall for a model tokamak equilibrium of KTX in the large aspect-ratio approximation. For the RFP configuration, the standard formulation of dispersion relation for RWM based on the MHD energy principle has been evaluated for a cylindrical α- Θ model of KTX plasma equilibrium, in an effort to investigate the effects of thin wall on the RWM in KTX. Full MHD calculations of RWM in the RFP configuration of KTX using the NIMROD code are also being developed. Supported by National Magnetic Confinement Fusion Science Program of China Grant Nos. 2014GB124002, 2015GB101004, 2011GB106000, and 2011GB106003.

  8. Fusion Energy-Production from a Deuterium-Tritium Plasma in the Jet Tokamak

    NARCIS (Netherlands)

    Rebut, P. H.; Gibson, A.; Huguet, M.; Adams, J. M.; Alper, B.; Altmann, H.; Andersen, A.; Andrew, P.; Angelone, M.; Aliarshad, S.; Baigger, P.; Bailey, W.; Balet, B.; Barabaschi, P.; Barker, P.; Barnsley, R.; Baronian, M.; Bartlett, D. V.; Baylor, L.; Bell, A. C.; Benali, G.; Bertoldi, P.; Bertolini, E.; Bhatnagar, V.; Bickley, A. J.; Binder, D.; Bindslev, H.; Bonicelli, T.; Booth, S. J.; Bosia, G.; Botman, M.; Boucher, D.; Boucquey, P.; Breger, P.; Brelen, H.; Brinkschulte, H.; Brooks, D.; Brown, A.; Brown, T.; Brusati, M.; Bryan, S.; Brzozowski, J.; Buchse, R.; Budd, T.; Bures, M.; Businaro, T.; Butcher, P.; Buttgereit, H.; Caldwellnichols, C.; Campbell, D. J.; Card, P.; Celentano, G.; Challis, C. D.; Chankin, A. V.; Cherubini, A.; Chiron, D.; Christiansen, J.; Chuilon, P.; Claesen, R.; Clement, S.; Clipsham, E.; Coad, J. P.; Coffey, I. H.; Colton, A.; Comiskey, M.; Conroy, S.; Cooke, M.; Cooper, D.; Cooper, S.; Cordey, J. G.; Core, W.; Corrigan, G.; Corti, S.; Costley, A. E.; Cottrell, G.; Cox, M.; Cripwell, P.; Dacosta, O.; Davies, J.; Davies, N.; de Blank, H.; De Esch, H.; Dekock, L.; Deksnis, E.; Delvart, F.; Dennehinnov, G. B.; Deschamps, G.; Dickson, W. J.; Dietz, K. J.; Dmitrenko, S. L.; Dmitrieva, M.; Dobbing, J.; Doglio, A.; Dolgetta, N.; Dorling, S. E.; Doyle, P. G.; Duchs, D. F.; Duquenoy, H.; Edwards, A.; Ehrenberg, J.; Ekedahl, A.; Elevant, T.; Erents, S.K.; Eriksson, L. G.; Fajemirokun, H.; Falter, H.; Freiling, J.; Freville, F.; Froger, C.; Froissard, P.; Fullard, K.; Gadeberg, M.; Galetsas, A.; Gallagher, T.; Gambier, D.; Garribba, M.; Gaze, P.; Giannella, R.; Gill, R. D.; Girard, A.; Gondhalekar, A.; Goodall, D.; Gormezano, C.; Gottardi, N. A.; Gowers, C.; Green, B. J.; Grievson, B.; Haange, R.; Haigh, A.; Hancock, C. J.; Harbour, P. J.; Hartrampf, T.; Hawkes, N. C.; Haynes, P.; Hemmerich, J. L.; Hender, T.; Hoekzema, J.; Holland, D.; Hone, M.; Horton, L.; How, J.; Huart, M.; Hughes, I.; Hughes, T. P.; Hugon, M.; Huo, Y.; Ida, K.; Ingram, B.; Irving, M.; Jacquinot, J.; Jaeckel, H.; Jaeger, J. F.; Janeschitz, G.; Jankovicz, Z.; Jarvis, O. N.; Jensen, F.; Jones, E. M.; Jones, H. D.; Jones, Lpdf; Jones, S.; Jones, T. T. C.; Junger, J. F.; Junique, F.; Kaye, A.; Keen, B. E.; Keilhacker, M.; Kelly, G. J.; Kerner, W.; Khudoleev, A.; Konig, R.; Konstantellos, A.; Kovanen, M.; Kramer, G.; Kupschus, P.; Lasser, R.; Last, J. R.; Laundy, B.; Laurotaroni, L.; Laveyry, M.; Lawson, K.; Lennholm, M.; Lingertat, J.; Litunovski, R. N.; Loarte, A.; Lobel, R.; Lomas, P.; Loughlin, M.; Lowry, C.; Lupo, J.; Maas, A. C.; Machuzak, J.; Macklin, B.; Maddison, G.; Maggi, C. F.; Magyar, G.; Mandl, W.; Marchese, V.; Marcon, G.; Marcus, F.; Mart, J.; Martin, D.; Martin, E.; Martinsolis, R.; Massmann, P.; Matthews, G.; McBryan, H.; McCracken, G.; McKivitt, J.; Meriguet, P.; Miele, P.; Miller, A.; Mills, J.; Mills, S. F.; Millward, P.; Milverton, P.; Minardi, E.; Mohanti, R.; Mondino, P. L.; Montgomery, D.; Montvai, A.; Morgan, P.; Morsi, H.; Muir, D.; Murphy, G.; Myrnas, R.; Nave, F.; Newbert, G.; Newman, M.; Nielsen, P.; Noll, P.; Obert, W.; Obrien, D.; Orchard, J.; Orourke, J.; Ostrom, R.; Ottaviani, M.; Pain, M.; Paoletti, F.; Papastergiou, S.; Parsons, W.; Pasini, D.; Patel, D.; Peacock, A.; Peacock, N.; Pearce, R. J. M.; Pearson, D.; Peng, J. F.; Desilva, R. P.; Perinic, G.; Perry, C.; Petrov, M.; Pick, M. A.; Plancoulaine, J.; Poffe, J. P.; Pohlchen, R.; Porcelli, F.; Porte, L.; Prentice, R.; Puppin, S.; Putvinskii, S.; Radford, G.; Raimondi, T.; Deandrade, M. C. R.; Reichle, R.; Reid, J.; Richards, S.; Righi, E.; Rimini, F.; Robinson, D.; Rolfe, A.; Ross, R. T.; Rossi, L.; Russ, R.; Rutter, P.; Sack, H. C.; Sadler, G.; Saibene, G.; Salanave, J. L.; Sanazzaro, G.; Santagiustina, A.; Sartori, R.; Sborchia, C.; Schild, P.; Schmid, M.; Schmidt, G.; Schunke, B.; Scott, S. M.; Serio, L.; Sibley, A.; Simonini, R.; Sips, A.C.C.; Smeulders, P.; Smith, R.; Stagg, R.; Stamp, M.; Stangeby, P.; Stankiewicz, R.; Start, D. F.; Steed, C. A.; Stork, D.; Stott, P.E.; Stubberfield, P.; Summers, D.; Summers, H.; Svensson, L.; Tagle, J. A.; Talbot, M.; Tanga, A.; Taroni, A.; Terella, C.; Terrington, A.; Tesini, A.; Thomas, P. R.; Thompson, E.; Thomsen, K.; Tibone, F.; Tiscornia, A.; Trevalion, P.; Tubbing, B.; Vanbelle, P.; Vanderbeken, H.; Vlases, G.; von Hellermann, M.; Wade, T.; Walker, C.; Walton, R.; Ward, D.; Watkins, M. L.; Watkins, N.; Watson, M. J.; Weber, S.; Wesson, J.; Wijnands, T. J.; Wilks, J.; Wilson, D.; Winkel, T.; Wolf, R.; Wong, D.; Woodward, C.; Wu, Y.; Wykes, M.; Young, D.; Young, I. D.; Zannelli, L.; Zolfaghari, A.; Zwingmann, W.

    1992-01-01

    The paper describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixtures. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to

  9. Physics of GAM-initiated L-H transition in a tokamak

    Science.gov (United States)

    Askinazi, L. G.; Belokurov, A. A.; Bulanin, V. V.; Gurchenko, A. D.; Gusakov, E. Z.; Kiviniemi, T. P.; Lebedev, S. V.; Kornev, V. A.; Korpilo, T.; Krikunov, S. V.; Leerink, S.; Machielsen, M.; Niskala, P.; Petrov, A. V.; Tukachinsky, A. S.; Yashin, A. Yu; Zhubr, N. A.

    2017-01-01

    Based on experimental observations using the TUMAN-3M and FT-2 tokamaks, and the results of gyrokinetic modeling of the interplay between turbulence and the geodesic acoustic mode (GAM) in these installations, a simple model is proposed for the analysis of the conditions required for L-H transition triggering by a burst of radial electric field oscillations in a tokamak. In the framework of this model, one-dimensional density evolution is considered to be governed by an anomalous diffusion coefficient dependent on radial electric field shear. The radial electric field is taken as the sum of the oscillating term and the quasi-stationary one determined by density and ion temperature gradients through a neoclassical formula. If the oscillating field parameters (amplitude, frequency, etc) are properly adjusted, a transport barrier forms at the plasma periphery and sustains after the oscillations are switched off, manifesting a transition into the high confinement mode with a strong inhomogeneous radial electric field and suppressed transport at the plasma edge. The electric field oscillation parameters required for L-H transition triggering are compared with the GAM parameters observed at the TUMAN-3M (in the discharges with ohmic L-H transition) and FT-2 tokamaks (where no clear L-H transition was observed). It is concluded based on this comparison that the GAM may act as a trigger for the L-H transition, provided that certain conditions for GAM oscillation and tokamak discharge are met.

  10. Feasibility of a multi-purpose demonstration neutron source based on a compact superconducting spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Guillemaut, C., E-mail: christophe.guillemaut@ccfe.ac.uk [Insituto de Ciencias Nucleares, Universidad Nacional Autónoma de México, A.P. 70-543, Ciudad Universitaria, 04511 Coyoacán, D.F. (Mexico); Herrera Velázquez, J.J.E. [Insituto de Ciencias Nucleares, Universidad Nacional Autónoma de México, A.P. 70-543, Ciudad Universitaria, 04511 Coyoacán, D.F. (Mexico); Suarez, A. [Laboratorio Nacional de Fusión, Asociación EURATOM-CIEMAT, 28040 Madrid (Spain)

    2013-12-15

    Tokamak neutron sources would allow near term applications of fusion such as fusion–fission hybrid reactors, elimination of nuclear wastes, production of radio-isotopes for nuclear medicine, material testing and tritium production. The generation of neutrons with fusion plasmas does not require energetic efficiency; thus, nowadays tokamak technologies would be sufficient for such purposes. This paper presents some key technical details of a compact (∼1.8 m{sup 3} of plasma) superconducting spherical tokamak neutron source (STNS), which aims to demonstrate the capabilities of such a device for the different possible applications already mentioned. The T-11 transport model was implemented in ASTRA for 1.5 D simulations of heat and particle transport in the STNS core plasma. According to the model predictions, total neutron production rates of the order of ∼10{sup 15} s{sup −1} and ∼10{sup 13} s{sup −1} can be achieved with deuterium/tritium and deuterium/deuterium respectively, with 9 MW of heating power, 1.4 T of toroidal magnetic field and 1.5 MA of plasma current. Engineering estimates indicate that such scenario could be maintained during ∼20 s and repeated every ∼5 min. The viability of most of tokamak neutron source applications could be demonstrated with a few of these cycles and around ∼100 cycles would be required in the worst cases.

  11. Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

    Science.gov (United States)

    Freidberg, J.; Mangiarotti, F.; Minervini, J.

    2014-10-01

    The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

  12. Scoping study for compact high-field superconducting net energy tokamaks

    Science.gov (United States)

    Mumgaard, R. T.; Greenwald, M.; Freidberg, J. P.; Wolfe, S. M.; Hartwig, Z. S.; Brunner, D.; Sorbom, B. N.; Whyte, D. G.

    2016-10-01

    The continued development and commercialization of high temperature superconductors (HTS) may enable the construction of compact, net-energy tokamaks. HTS, in contrast to present generation low temperature superconductors, offers improved performance in high magnetic fields, higher current density, stronger materials, higher temperature operation, and simplified assembly. Using HTS along with community-consensus confinement physics (H98 =1) may make it possible to achieve net-energy (Q>1) or burning plasma conditions (Q>5) in DIII-D or ASDEX-U sized, conventional aspect ratio tokamaks. It is shown that, by operating at high plasma current and density enabled by the high magnetic field (B>10T), the required triple products may be achieved at plasma volumes under 20m3, major radii under 2m, with external heating powers under 40MW. This is at the scale of existing devices operated by laboratories, universities and companies. The trade-offs in the core heating, divertor heat exhaust, sustainment, stability, and proximity to known plasma physics limits are discussed in the context of the present tokamak experience base and the requirements for future devices. The resulting HTS-based design space is compared and contrasted to previous studies on high-field copper experiments with similar missions. The physics exploration conducted with such HTS devices could decrease the real and perceived risks of ITER exploitation, and aid in quickly developing commercially-applicable tokamak pilot plants and reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  14. ELM induced tungsten melting and its impact on tokamak operation

    Science.gov (United States)

    Coenen, J. W.; Arnoux, G.; Bazylev, B.; Matthews, G. F.; Jachmich, S.; Balboa, I.; Clever, M.; Dejarnac, R.; Coffey, I.; Corre, Y.; Devaux, S.; Frassinetti, L.; Gauthier, E.; Horacek, J.; Knaup, M.; Komm, M.; Krieger, K.; Marsen, S.; Meigs, A.; Mertens, Ph.; Pitts, R. A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.

    2015-08-01

    In JET-ILW dedicated melt exposures were performed using a sequence of 3MA/2.9T H-Mode JET pulses with an input power of PIN = 23 MW, a stored energy of ∼6 MJ and regular type I ELMs at ΔWELM = 0.3 MJ and fELM ∼ 30 Hz. In order to assess the risk of starting ITER operations with a full W divertor, one of the task was to measure the consequences of W transients melting due to ELMs. JET is the only tokamak able to produce transients/ ELMs large enough (>300 kJ per ELM) to facilitate melting of tungsten. Such ELMs are comparable to mitigated ELMs expected in ITER. By moving the outer strike point (OSP) onto a dedicated leading edge the base temperature was raised within ∼1 s to allow transient ELM-driven melting during the subsequent 0.5 s. Almost 1 mm (∼6 mm3) of W was moved by ∼ 150 ELMs within 5 subsequent discharges. Significant material losses in terms of ejections into the plasma were not observed. There is indirect evidence that some small droplets (∼ 80 μm) were ejected. The impact on the main plasma parameters is minor and no disruptions occurred. The W-melt gradually moved along the lamella edge towards the high field side, driven by j × B forces. The evaporation rate determined is 100 times less than expected from steady state melting and thus only consistent with transient melting during individual ELMs. IR data, spectroscopy, as well as melt modeling point to transient melting. Although the type of damage studied in these JET experiments is unlikely to be experienced in ITER, the results do strongly support the design strategy to avoid exposed edges in the ITER divertor. The JET experiments required a surface at normal incidence and considerable pre-heating to produce tungsten melting. They provide unique experimental evidence for the absence of significant melt splashing at events resembling mitigated ELMs on ITER and establish a unique experimental benchmark for the simulations being used to study transient shallow melting on ITER W

  15. Noble magnetic barriers in the ASDEX UG tokamak

    Science.gov (United States)

    Ali, Halima; Punjabi, Alkesh; Vazquez, Justin

    2010-02-01

    The second-order perturbation method of creating invariant tori inside chaos in Hamiltonian systems (Ali, H.; Punjabi, A. Plasma Phys. Contr. F. 2007, 49, 1565-1582) is applied to the axially symmetric divertor experiment upgrade (ASDEX UG) tokamak to build noble irrational magnetic barriers inside chaos created by resonant magnetic perturbations (m, n)=(3, 2)+(4, 3), with m and n the poloidal and toroidal mode numbers of the Fourier expansion of the magnetic perturbation. The radial dependence of the Fourier modes is ignored. The modes are considered to be locked and have the same amplitude δ. A symplectic mathematical mapping in magnetic coordinates is used to integrate magnetic field line trajectories in the ASDEX UG. Tori with noble irrational rotational transform are the last ones to be destroyed by perturbation in Hamiltonian systems. For this reason, noble irrational magnetic barriers are built inside chaos, and the strongest noble irrational barrier is identified. Three candidate locations for the strongest noble barrier in ASDEX UG are selected. All three candidate locations are chosen to be roughly midway between the resonant rational surfaces ψ32 and ψ43. ψ is the magnetic coordinate of the flux surface. The three candidate surfaces are the noble irrational surfaces close to the surface with q value that is a mediant of q=3/2 and 4/3, q value of the physical midpoint of the two resonant surfaces, and the q value of the surface where the islands of the two perturbing modes just overlap. These q values of the candidate surfaces are denoted by q MED, q MID, and q OVERLAP. The strongest noble barrier close to q MED has the continued fraction representation (CFR) [1;2,2,1∞] and exists for δ≤2.6599×10-4; the strongest noble barrier close to q MID has CFR [1;2,2,2,1∞] and exists for δ≤4.6311×10-4; and the strongest noble barrier close to q OVERLAP has CFR [1;2,2,6,2,1∞] and exists for δ≤1.367770×10-4. From these results, the strongest

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

    Science.gov (United States)

    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

  17. Online impedance matching system for ICRH-RF experiments on SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, R.; Singh, M.; Jadav, H.M.; Mishra, K.; Singh, Raj; Kulkarni, S.V.; Bora, D.

    2015-11-15

    High-power radio frequency (RF) heating system in the frequency range of ion cyclotron range of frequencies (ICRF) on tokamak needs a mechanism for matching impedance seen by the antenna in presence of plasma to the generator output impedance for maximum transfer of RF power to the plasma with minimum reflections to avoid damage to the generator. The impedance of the antenna is strongly dependent on edge plasma parameters and the impedance sometimes changes as fast as 10{sup −4} s, while the RF generators used can deliver full power only into constant load impedance. Hence, the matching system with dynamic response in between generator and antenna is very much essential for high power ICRF experiments when the plasma is of longer duration with variable load impedance. For ICRF system on SST-1 tokamak, two automatic matching systems are employed each for one transmission line, which consists of motorized stub tuners and phase shifters to match the antenna impedance in the time scale of 120 ms. As a part of initial testing, online impedance matching system is tested with individual transmission lines and then both the lines are matched simultaneously on a variable dummy load which simulates the plasma load. In order to deliver power to both the lines from a single RF generator, hybrid coupler is used which also protects RF generator from reflections up to certain extent. However for hybrid coupler has to work properly and both the lines should see same load impedance. The automatic matching system is installed on tokamak state super-conducting tokamak (SST-1) and is tested up to 140 kW power in the vacuum vessel of the tokamak. Here we present the details of the online matching system and its testing results. The significant result is that we could match the variable load impedance with the generator impedance within 120 ms by suitably adjusting the step counts of the motor controller.

  18. Status of design and experimental activity on module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, Igor E., E-mail: lyublinski@yandex.ru [JSC “Red Star”, Moscow (Russian Federation); Vertkov, Alexey V.; Zharkov, Mikhail Yu.; Semenov, Vladimir V. [JSC “Red Star”, Moscow (Russian Federation); Mirnov, Sergey V.; Lazarev, Vladimir B. [GSC RF TRINITI, Troitsk, Moscow Region (Russian Federation); Tazhibayeva, Irina L.; Shapovalov, Gennadiy V.; Kulsartov, Timur V.; D’yachenko, Alexandr V. [IAE of National Nuclear Center, Kurchatov (Kazakhstan); Mazzitelli, Giuseppe [Associazione EURATOM-ENEA sulla Fusione, C.R. ENEA Frascati, Rome (Italy); Agostini, Pietro [ENEA Brasimone, Camugnano, BO (Italy)

    2013-10-15

    Highlights: • Lithium divertor module based on capillary-porous system is created for KTM tokamak. • The hydraulic tests of lithium divertor module were conducted. • The results were compared with the calculation data. • The analysis of results’ discrepancies was conducted. • The lithium divertor module is ready for tests on KTM tokamak. -- Abstract: The projects of ITER and DEMO reactors showed that there are serious difficulties with solving the issues of plasma facing elements (PFE) based on the solid materials. Problems of PFE can be overcome by the use of liquid lithium. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material, in which liquid lithium fills a solid matrix from porous material. The progress in development of lithium technology and also lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, LTX, HT-7 and stellarator TJ II is a good basis for development of the project of steady-state operating lithium divertor module for Kazakhstan tokamak. At present the lithium divertor module for KTM tokamak is development and manufacturing. The paper describes main design features of the module of lithium divertor (MLD). The first step of the hydraulic tests of MLD with fully assembled external thermo-stabilization system, which was connected to in-vessel lithium unit, were performed using ethanol as a model heat transfer media. Test results of MLD have shown that operating parameters of designed and manufactured system for thermo-stabilization are sufficient for proper operation; basic hydraulic characteristics of the system are close to expected values.

  19. Virtual reality applications in remote handling development for tokamaks in India

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, Pramit, E-mail: pramitd@ipr.res.in; Rastogi, Naveen; Gotewal, Krishan Kumar

    2017-05-15

    Highlights: • Evaluation of Virtual Reality (VR) in design and operation phases of Remote Handling (RH) equipment for tokamak. • VR based centralized facility, to cater RH development and operation, is setup at Institute for Plasma Research, India. • The VR facility system architecture and components are discussed. • Introduction to various VR applications developed for design and development of tokamak RH equipment. - Abstract: A tokamak is a plasma confinement device that can be used to achieve magnetically confined nuclear fusion within a reactor. Owing to the harsh environment, Remote Handling (RH) systems are used for inspection and maintenance of the tokamak in-vessel components. As the number of in-vessel components requiring RH maintenance is large, physical prototyping of all strategies becomes a major challenge. The operation of RH systems poses further challenge as all equipment have to be controlled remotely within very strict accuracy limits with minimum reliance on the available camera feedback. In both design and operation phases of RH equipment, application of Virtual Reality (VR) becomes imperative. The scope of this paper is to introduce some applications of VR in the design and operation cycle of RH, which are not available commercially. The paper discusses the requirement of VR as a tool for RH equipment design and operation. The details of a comprehensive VR facility that has been established to support the RH development for Indian tokamaks are also presented. Further, various cases studies are provided to highlight the utilization of this VR facility within phases of RH development and operation.

  20. Comparative study of runaway electron diffusion in the rise phase of low and normal discharges in the SINP tokamak

    Indian Academy of Sciences (India)

    Ramesh Narayanan; A N Sekar Iyengar

    2010-10-01

    The behaviour of runaway electrons in the SINP tokamak, which can be operated in a normal edge safety factor () (NQ) discharge configuration as well as in a low (LQ) configuration, was experimentally investigated, during the initial plasma generation phase. An energy analysis of the runaway electron dynamics in the rise phase of the SINP tokamak discharges was also made. A comparison of the runaway electron diffusion coefficients in NQ and LQ is carried out in this paper.