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Sample records for hl-2a tokamak plasmas

  1. Initial analysis of the HL-2A tokamak visible light measurements

    International Nuclear Information System (INIS)

    Shi Le; Zheng Yinjia

    2005-01-01

    Measurement of H α radial distribution is an important part in a plasma experiment. During the HL-2A experiment, CCD photos recorded the plasma radiation information using a plasma imaging system, particularly the H α line emission information (with H α filter in front of the camera). Since the line of light radiation measurement involves a line integral of the radiating region, data handling and analysis are necessary. The H α radial distribution along an HL-2A tokamak radial radius can be obtained by using an Abel inversion transformation with plasma imaging on the HL-2A tokamak. This paper provides a reliable measurement method on the HL-2A tokamak. (authors)

  2. Physics design of the HL-2A tokamak

    International Nuclear Information System (INIS)

    Gao Qingdi; Li Fangzhu; Zhang Jinhua; Pan Yudong; Jiao Yiming

    2005-10-01

    An overview report for the physics design of the HL-2A tokamak is presented. By numerically analyzing the plasma shaping and the vertical instability due to plasma elongation, the requirements for the currents of poloidal magnetic field coils and the control system are put forward. Controlling the plasma profile by using NBI (neutral beam injection) and LHCD (lower hybrid current drive) is investigated, and the high performance modes of operation in HL-2A and modeled and designed. The magnetohydrodynamic instabilities in improved confinement configuration (RS configuration) are studied so as to point out the way of plasma control to perform stationary high performance discharges in HL-2A. In order to offer data for updating the HL-2A divertor, performances of the divertor plasma are simulated. (authors)

  3. Recent advances in the HL-2A tokamak experiments

    International Nuclear Information System (INIS)

    Liu, Y.; Ding, X.T.; Yang, Q.W.; Yan, L.W.; Liu, D.Q.; Xuan, W.M.; Chen, L.Y.; Song, X.M.; Cao, Z.; Zhang, J.H.; Mao, W.C.; Zhou, C.P.; Li, X.D.; Wang, S.J.; Yan, J.C.; Bu, M.N.; Chen, Y.H.; Cui, C.H.; Cui, Z.Y.; Deng, Z.C.; Hong, W.Y.; Hu, H.T.; Huang, Y.; Kang, Z.H.; Li, B.; Li, W.; Li, F.Z.; Li, G.S.; Li, H.J.; Li, Q.; Li, Y.G.; Li, Z.J.; Liu, Yi; Liu, Z.T.; Luo, C.W.; Mao, X.H.; Pan, Y.D.; Rao, J.; Shao, K.; Song, X.Y.; Wang, M.; Wang, M.X.; Wang, Q.M.; Xiao, Z.G.; Xie, Y.F.; Yao, L.H.; Yao, L.Y.; Zheng, Y.J.; Zhong, G.W.; Zhou, Y.; Pan, C.H.

    2005-01-01

    Two experiment campaigns were conducted on the HL-2A tokamak in 2003 and 2004 after the first plasma was obtained at the end of 2002. Progresses in many aspects have been made, especially in the divertor discharge and feedback control of plasma configuration. Up to now, the following operation parameters have been achieved: I p = 320 kA, B t = 2.2 T and discharge duration T d = 1580 ms. With the feedback control of plasma current and horizontal position, an excellent repeatability of the discharge has been achieved. The tokamak has been operated at both limiter configuration and single null (SN) divertor configuration. The HL-2A SN divertor configuration is simulated with the MHD equilibrium code SWEQU. When the divertor configuration is formed, the impurity radiation in the main plasma decreases remarkably

  4. On the HL-1M tokamak plasma confinement time

    International Nuclear Information System (INIS)

    Qin Yunwen

    2001-01-01

    Emphasizing that the tokamak plasma confinement time is the plasma particle or thermal energy loss characteristic time, the relevant physical concept and HL-1M tokamak experimental data analyses are reviewed

  5. The conceptual design for the modification of HL-2A tokamak

    International Nuclear Information System (INIS)

    Dequan Liu

    2006-01-01

    The medium-sized tokamak HL-2A, based on the former ASDEX's main components has approached its rating operational parameters with I p = 450 kA and B T =2.8 T so far. The HL-2A was originally designed to operate under an axisymmetrical double-null configuration by two triplets of shaping coils located in the vacuum vessel. Because the shaping coils were set very close to the plasma column, the nulls are normally fixed and the plasma has a nearly circular cross-section. It is difficult to increase the parameters further and obtain a preferable plasma shape with certain values of triangularity and elongation. Due to the defects on upper shaping coils' electrical insulation, the HL-2A is presently just operated with lower single null divertor configuration. Additionally, the experiment has to occasionally terminate due to incidental damages of the weak parts. With the experimental progress on HL-2A, the plasma stored energy and auxiliary heating power is being increased, a more efficient, compact and tight divertor is needed. Based on present status of HL-2A tokamak and for satisfying the requirements of the experiment, the modification of HL-2A tokamak device will be carried out. The modification design of HL-2A intents to obtain optimized plasma parameters, e.g., aspect ratio, I p , b, plasma volume, certain plasma shaping with preferable elongation and triangularity, a stable vacuum vessel easy to be maintained and an advanced divertor. The primary consideration of the modification is to take all of the shaping coils out of the vessel to enlarge the plasma volume. The elongated and triangular cross section with double null of plasma is to be obtained by composite distribution of eight sets of poloidal coils. Because all poloidal coils will be located between the TF coil and the vessel, the vacuum vessel will be remade with a smaller size to make rooms for the poloidal coil system. With an optimized operation mode, the volt-second of the plasma may be increased a

  6. Commissioning and preliminary operation of HL-2A tokamak

    International Nuclear Information System (INIS)

    Liu Dequan; Liu Yong; Yan Jianchen; Cao Zeng; Yang Qingwei; Zhou Caipin; Li Xiaodong

    2005-01-01

    HL-2A is a divertor tokamak located at new site of Southwestern Institute of Physics (SWIP), Chengdu, China. It can be operated in double-null and single-null divertor with closed configurations. The effects of the divertor on impurity behaviors, MHD instabilities, transport, wall conditioning and divertor physics are key issues to be studied during the first step operation on HL-2A. Preliminary experiments with limiter and single null divertor configurations were carried out in 2003. Main parameters achieved are the plasma current 168 KA, duration time of 920 ms and plasma linear-averaged density of 1.7 x 10 19 m -3 . The impurities, especially of low Z, are clearly decreased during the divertor experiments

  7. Feedback control of plasma position in the HL-1 tokamak

    International Nuclear Information System (INIS)

    Yuan Baoshan; Jiao Boliang; Yang Kailing

    1991-01-01

    In the HL-1 tokamak with a thick copper shell, the control of plasma position is successfully performed by a feedback-feedforward system with dual mode regulator and the equilibrium field coils outside the shell. The plasma position can be controlled within ±2 mm in both vertical and horizontal directions under the condition that the iron core of transformer is not saturated

  8. Study on fast ion loss in HL-2A tokamak

    International Nuclear Information System (INIS)

    Liu Yi; Sun Tengfei; Ji Xiaoquan

    2012-01-01

    Experiments with a high-energy deuterium neutral beam (NB) injection (30 keV, about 0.6 MW) were performed on the HL-2A tokamak. Analysis of neutron decay following the NB 'blip' injection indicates that tangentially injected beam ions are well confined, slowing down classically in the HL-2A. Anomalous losses of beam ions were observed when a beta-induced Alfven acoustic (BAAE) mode was present in the plasma. Such a high energetic particle driven mode led to fast-ion loss, showing a strong influence of the energetic particle driven mode on the fast-ion transport. (authors)

  9. Influence of large dust particles on plasma performance in the HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Z.H., E-mail: huangzh@swip.ac.cn; Yan, L.W.; Feng, Z.; Cheng, J.; Tomita, Y.; Liu, L.; Gao, J.M.; Zhong, W.L.; Jiang, M.; Yang, Q.W.; Xu, Y.; Duan, X.R.

    2015-08-15

    Visible dust particles generated from plasma-facing components (PFCs) and the impact of the dusts on plasma performance as a source of impurities have been studied in the HL-2A tokamak by means of a fast framing camera together with other diagnostics. The camera images display that during a steady state discharge the dusts are accelerated toriodally by the ion drag force and radially by the centrifugal force. The first experimental evidence shows that dust particles originating from the high field side (HFS) lead to a significant reduction of central electron temperature and divertor heat flux, a considerable rise of total radiated power and effective charge, and a slight growth of local electron density. The results reveal that the dusts at the HFS have much stronger effects on plasma performance than those at the low field side (LFS)

  10. Investigation of plasma equilibrium in HL-1 tokamak

    International Nuclear Information System (INIS)

    Lu Zhihong; Jiang Yunxia; Yang Jinwei; Zhang Baozhu; Qiu Wei; Qin Yunwen

    1987-01-01

    In this paper, the plasma equilibrium in HL-1 tokamak has been discussed. The horizontal and vertical displacement of plasma is measured using a symmetical magneic probe system, and the temporal evolution of displacements is given by a data acquisition system with micro-computer. The influence of various stray fields on plasma equilibrium has been analysed. The direction and value of horizontal stray field induced by the totoidal field coils and primary windings are determined using vertical displacement data. By adjusting parameters of internal and external vertical field, in the flat part of discharge current, the plasma can be kept in its equilibuium state at the place where is 3 cm outer of chamber certre, i.e nearby the centre of limiter

  11. Preliminary results of MHD stability in HL-1 tokamak

    International Nuclear Information System (INIS)

    Zheng Yongzhen; Ma Tengcai; Xiao Zhenggui Cai Renfang

    1987-01-01

    In this paper, MHD activities of HL-1 tokamak plasma are studied with Fourier transform and correlatio analysis. The poloidal modes m = 1, 2, 3,4 and toroidal modes n of MHD magnetic fluctuation signals are detected. Methods for suppressing MHD instabilities are suggested and tested, after MHD instabilities are studied in HL-1. The effects of MHD characteristics in the beginning stage of discharge on the whole process of discharge are analyzed. The disruption, in HL-1 device could be divided into three kinds: internal disruption, minor disruption and major disruption. The result shows that HL-1 will have a better operation condition if internal disruption appears. In is end, the stable operation region of HL-1 tokamak is also given

  12. Plasma electron density measurement with multichannel microwave interferometer on the HL-1 tokamak device

    International Nuclear Information System (INIS)

    Xu Deming; Zhang Hongyin; Liu Zetian; Ding Xuantong; Li Qirui; Wen Yangxi

    1989-11-01

    A multichannel microwave interferometer which is composed of different microwave interferometers (one 2 mm band, one 4 mm band and two 8 mm band) has been used to measure the plasma electron density on HL-1 tokamak device. The electron density approaching to 5 x 10 13 cm -3 is measured by a 2 mm band microwave interferometer. In the determinable range, the electron density profile in the cross-section on HL-1 device has been measured by this interferometer. A microcomputer data processing system is also developed

  13. A discussion on practicable scheme for machine control system of HL-2A tokamak

    International Nuclear Information System (INIS)

    Li Qiang; Song Xianming; Jiang Chao

    2001-01-01

    This is the modified version of The Preliminary Design on the Hardware and Nets of HL-2A Tokamak. In this version, centralized control as well as the field bus communication on HL-2A Tokamak is used. The hardware components and the operational theory are introduced. The questions of practice program, extensibility, centralized control, cooperation with other subsystems and the continuous adaptation of present device are all discussed. The budget of the system is detailed. To keep the step with the overall engineering constructions of HL-2A, suggestions of the time program are presented for the system design, instrument purchases, installation, construction, user program development and the final operation processes for the machine control system of HL-2A Tokamak

  14. Simulation of the MHD stabilities of the experiment on HL-2A tokamak by GATO code

    International Nuclear Information System (INIS)

    Pan Wei; Chen Liaoyuan; Dong Jiaqi; Shen Yong; Zhang Jinhua

    2009-01-01

    The ideal two-dimensional MHD stabilities code, GATO, has been successfully immigrated to the high-performance computing system of HL-2A and used to the simulation study of the ideal MHD stabilities of the plasmas produced by one of the pellets injection experiments on HL-2A tokamak. The EFIT code was used to reconstruct the equilibrium configures firstly and the GATO was used to compute their MHD stabilities secondly whose source data were obtained by the NO.4050 discharge of the experiments on HL-2A, and finally by analyzing these results the preliminary conclusion was devised that the confinement performance of the plasma was improved because of the stabilization effect of the anti-sheared configures created by the pellets injection. (authors)

  15. Fast reciprocating probe system on the HL-2A tokamak

    International Nuclear Information System (INIS)

    Yan Longwen; Hong Wenyu; Qian Jun; Luo Cuiwen; Pan Li

    2005-01-01

    A reciprocating probe system has been installed at the midplane of the HL-2A tokamak. The probe is used to measure plasma edge density, temperature, floating potential, and corresponding fluctuation profiles with 8 cm scan from the scrape-off layer to the plasma boundary. The reciprocating probe can move at a speed of 1 m/s. A digital grating displacement measurement system that can provide a high displacement resolution of 0.04 mm is applied to the reciprocating probe system for the first time. A port located behind the vacuum isolation valve is designed for viewing and the exchange of the probe head. Different probe heads can be used to satisfy different experimental requirements. The first probe head had four graphite measurement tips. For high frequency response, no isolation amplifier is used in the electric circuit of the probe measurement. A personal computer via an analog-to-digital digitizer card acquires probe system data, which are sent to a data server by optical fiber after a discharge. All data are sent to the centralized data management system of the HL-2A. In this article we presented the edge temperature and density profiles for the limiter and divertor configurations of a selected plasma discharge

  16. Plasma behavior with molecular beam injection in the HL-1m tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Tang Nianyi; Cui Zhengying; Xu Deming; Deng Zhongchao; Ding Xuantong; Luo Junlin; Dong Jiafu; Guo Gancheng; Yang Shikun; Cui Chenghe; Xiao Zhenggui; Liu Dequan; Chen Xiaoping; Yan Longwen; Yan Donghai; Wang Enyao; Deng Xiwen

    1999-01-01

    The authors report effect of the new fueling method of high speed molecular beam injection on Tokamak confinement improvement. The present method is an improvement of conventional gas puffing, with performance comparable to the small pellet injection in HL-1M and also to the slow pellet in ASDEX. The fact that a shallower fueling can lead to similar confinement improvement as a deep one suggests that there may exist a critical position in a Tokamak plasma such that any kind of fueling will have a better confinement as long as it can give rise to density peaking at the critical position

  17. Conceptual design of HL-2M tokamak control system

    International Nuclear Information System (INIS)

    Xia Fan; Chen Liaoyuan; Song Xianming; Zhang Jinhua; Lou Cuiwen; Pan Yudong

    2009-01-01

    The static architecture, dynamic behavior, control theory and simulation of HL-2M tokamak control system are described. The real-time network will be build for the communication of real-time control among its subsystems and universal timing system will be build to guarantee the synchronization among the subsystems. The duty to achieve preprogrammed parameters is carried out by plasma discharge control. In order to reduce the damage made by discharge exception, the error-handing mechanism of supervision system is considered. The controllers of magnetic control system are designed to control the current, shape and position of plasma and simulation system is designed for testing the controllers. (authors)

  18. High spatial and temporal resolution charge exchange recombination spectroscopy on the HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Y. L.; Yu, D. L., E-mail: yudl@swip.ac.cn; Liu, L.; Cao, J. Y.; Sun, A. P.; Ma, Q.; Chen, W. J.; Liu, Yi; Yan, L. W.; Yang, Q. W.; Duan, X. R.; Liu, Yong [Southwestern Institute of Physics, Chengdu 610041 (China); Ida, K. [National Institute for Fusion Science, Toki 509-5292 (Japan); Hellermann, M. von [ITER Diagnostic Team, IO, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); FOM-Institute for Plasma physics “Rijnhuizen,” Association EURATOM, Trilateral Euregio Cluster, 3430 BE Nieuwegein (Netherlands)

    2014-10-01

    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 enable the temporal resolution of the system up to 400 Hz. The observation geometry and an optimized optic system enable the highest radial resolution up to ~1 cm at the plasma edge. The CXRS system monitors the carbon line emission (C VI, n = 8–7, 529.06 nm) whose Doppler broadening and Doppler shift provide ion temperature and plasma rotation velocity during the neutral beam injection. The composite CX spectral data are analyzed by the atomic data and analysis structure charge exchange spectroscopy fitting (ADAS CXSFIT) code. First experimental results are shown for the case of HL-2A plasmas with sawtooth oscillations, electron cyclotron resonance heating, and edge transport barrier during the high-confinement mode (H-mode)

  19. Mitigation and prediction of disruption on the HL-2A Tokamak

    International Nuclear Information System (INIS)

    Yong-Zhen, Zheng; Ying, Qiu; Peng, Zhang; Yuan, Huang; Zheng-Ying, Cui; Ping, Sun; Qing-Wei, Yang

    2009-01-01

    Injection of high-Z impurities into plasma has been proved to be able to reduce the localized thermal load and mechanical forces on the in-vessel components and the vacuum vessel, caused by disruptions in Tokamaks. An advanced prediction and mitigation system of disruption is implemented in HL-2A to safely shut down plasmas by using the laser ablation of high-Z impurities with a perturbation real-time measuring and processing system. The injection is usually triggered by the amplitude and frequency of the MHD perturbation field which is detected with a Mirnov coil and leads to the onset of a mitigated disruption within a few milliseconds. It could be a simple and potential approach to significantly reducing the plasma thermal energy and magnetic energy before a disruption, thereby achieving safe plasma termination. The plasma response to impurity injection, a mechanism for improving plasma thermal and current quench in major disruptions, the design of the disruption prediction warner, and an evaluation of the mitigation success rate are discussed in the present paper. (fluids, plasmas and electric discharges)

  20. Characterized the pattern of the material deposition in the HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Cai, Laizhong, E-mail: cailz@swip.ac.cn [Southwestern Institute of Physics, P.O.Box 432, Chengdu, Sichuan 610041 (China); Wang, Jianbao; Wu, Ting; Zeng, Xiaoxiao [Southwestern Institute of Physics, P.O.Box 432, Chengdu, Sichuan 610041 (China); Hai, Ran; Ding, Hongbin [Dalian University of Technology, Dalian, Liaoning 116024 (China)

    2017-03-15

    Since the divertor geometry of a tokamak has a strong impact on the material erosion and deposition on the wall and HL-2A has a unique divertor configuration, it is necessary to investigate the material deposition pattern in HL-2A although a few results on other tokamaks have already been published. In this paper, tiles retrieved from the vessel are analyzed ex-situ by SIMS, SEM and laser-induced breakdown spectroscopy (LIBS). And deposition behind the lower divertor is in-situ measured by a quartz crystal microbalance (QMB). The deposition in HL-2A displays a complex pattern and clear localization characteristic. The thickness of the deposition layer varies in the range of 0-4μm. And in-situ diagnostic of QMB indicates that the average thickness of the deposition layer per pulse is over ten nanometers. In addition, the results imply that Si, Fe and D have different behaviors during the material deposition in HL-2A.

  1. Development of plasma current waveform adjusting system ZLJ for tokamak device HL-1

    International Nuclear Information System (INIS)

    Wang Shangbing; Hu Haotian; Tang Fangqun; Zhou Yongzheng; Chu Xiuzhong; Cheng Jiashun; Gao Yunxia

    1989-12-01

    The control of some typical Tokamak discharge waveforms has been achieved by using plasma current waveform adjusting system ZLJ in the ohmic heating of HL-1. The discharge waveforms include a series of regular plasma current waveforms with various slow rising rate, such as 80 kA, 450 ms long flat-topping; 100 kA, 200 ms rising; 200 ms falt-topping and 180 kA, 400 ms slow rising etc. The design principle of the system and the initial experimental results are described

  2. Experimental progress and innovation on the HL-2A tokamak

    International Nuclear Information System (INIS)

    Ding Xuantong

    2010-01-01

    The HL-2A Tokamak is the first large controlled fusion experiment device with divertor in China. In this paper, the main experimental results on this device will be presented. Since its establishment, the operation conditions have been improved greatly. First, the divertor configuration was realized, the electron temperature was increased to 5.5 keV by electron cyclotron heating, and then the high confinement edge localized mode was achieved. With the development of high power auxiliary heating and advanced plasma diagnostic systems,innovative contributions have been made in certain plasma physics areas. The 3-dimension structure of the zonal flow has been identified, which is very important in the transport of fusion plasma. Supersonic molecular beam injection has also been developed and successfully used for plasma transport. The tearing mode has been suppressed by electron cyclotron resonance heating with low frequency modulation, and the confinement has been improved. The new phenomena of the internal kink mode and Alfen mode excited by energetic electrons have been observed. Future plans and new experiments on the device will also be briefly presented. (authors)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  4. Synchronous oscillation prior to disruption caused by kink modes in HL-2A tokamak plasmas

    Science.gov (United States)

    Jiang, M.; Hu, D.; Wang, X. G.; Shi, Z. B.; Xu, Y.; Chen, W.; Ding, X. T.; Zhong, W. L.; Dong, Y. B.; Ji, X. Q.; Zhang, Y. P.; Gao, J. M.; Li, J. X.; Yang, Z. C.; Li, Y. G.; Liu, Y.

    2015-08-01

    A class of evident MHD activities prior to major disruption has been observed during recent radiation induced disruptions of the HL-2A tokamak discharges. It can be named SOD, synchronous oscillations prior to disruption, characterized by synchronous oscillation of electron cyclotron emission (ECE), core soft x-ray, Mirnov coil, and {{D}α} radiation signals at the divertor plate. The SOD activity is mostly observed in a parametric regime where the poloidal beta is low enough before disruption, typically corresponding to those radiation-induced disruptions. It has been found that the m/n = 2/1 mode is dominant during the SODs, and consequently it is the drop of the mode frequency and the final mode locking that lead to thermal quench. The mode frequency before the mode locking corresponds to the toroidal rotation frequency of the edge plasma. It is also found that during SODs, the location of the q = 2 surface is moving outward, and most of the plasma current is enclosed within the surface. This demonstrates that the current channel lies inside the rational surface during SOD, and thus the resistive kink mode is unstable. Further analysis of the electron temperature perturbation structure shows that the plasma is indeed dominated by the resistive kink mode, with kink-like perturbation in the core plasma region. It suggests that it is the nonlinear growth of the m/n = 2/1 resistive kink mode and its higher order harmonics, rather than the spontaneous overlapping of multiple neighboring islands, that ultimately triggered the disruption.

  5. Analysis of the plasma magnetohydrodynamic equilibrium in iron core transformer Tokamak HL-1M

    International Nuclear Information System (INIS)

    Chen Xiaoguang; Yuan Baoshan

    1992-01-01

    The physical and mathematical model are presented on the problem of MHD equilibrium with the self consistent in iron core transformer HL-1M. Calculation and analysis for the plasma equilibrium of the stable boundary and free boundary are shown respectively, in an axisymmetric equilibrium model of two dimensions. First, a variation formulation of the problem is written and the equations of the poloided flux ψ are solved by a finite element method; the Picard and Newton algorithms are tested for the non-linearities. The plasma boundary and the magnetic surfaces are being simulated, with the currents in the coils, the total plasma current, its current density function and the magnetic permeability of the iron being the data for the problem; a certain number of the characteristic parameter of the equilibrium configuration is calculated. Secondly, a simple method of calculation is adopted in the determination of equilibrium fields and currents in iron core HL-1M tokamak device. In the plasma equilibrium, the magnetic effect of the air gaps in the iron core and the iron magnetic shielded plate are considered in HL-1M device. Reliable data are provided for designing and constructing the poloidal field system of HL-1M device. A good computer code is constructed, which may be useful in operating on analysis for the future device

  6. Control system of neoclassical tearing modes in real time on HL-2A tokamak

    Science.gov (United States)

    Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

    2017-11-01

    The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

  7. Preliminary Study of Ideal Operational MHD Beta Limit in HL-2A Tokamak Plasmas

    International Nuclear Information System (INIS)

    Shen Yong; Dong Jiaqi; He Hongda; Turnbull, A. D.

    2009-01-01

    Magnetohydrodynamic (MHD) n = 1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q 0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of β c N = 2.56 for a typical HL-2A discharge with a plasma current I p = 0.245 MA, and the scaling of β c N ∼constant is confirmed. (magnetically confined plasma)

  8. Empirical Scaling Laws of Neutral Beam Injection Power in HL-2A Tokamak

    International Nuclear Information System (INIS)

    Cao Jian-Yong; Wei Hui-Ling; Liu He; Yang Xian-Fu; Zou Gui-Qing; Yu Li-Ming; Li Qing; Luo Cui-Wen; Pan Yu-Dong; Jiang Shao-Feng; Lei Guang-Jiu; Li Bo; Rao Jun; Duan Xu-Ru

    2015-01-01

    We present an experimental method to obtain neutral beam injection (NBI) power scaling laws with operating parameters of the NBI system on HL-2A, including the beam divergence angle, the beam power transmission efficiency, the neutralization efficiency and so on. With the empirical scaling laws, the estimating power can be obtained in every shot of experiment on time, therefore the important parameters such as the energy confinement time can be obtained precisely. The simulation results by the tokamak simulation code (TSC) show that the evolution of the plasma parameters is in good agreement with the experimental results by using the NBI power from the empirical scaling law. (paper)

  9. The Study of Sawtooth Oscillation during ECRH of HL-2A-like Plasma using 1.5D BALDUR Code

    International Nuclear Information System (INIS)

    Promping, J.; Onjun, T.; Poolyarat, N.; Picha, R.

    2009-07-01

    Full text: One of the current issues in tokamak plasma is sawtooth oscillation, because each sawtooth crash results in a significant reduction of central temperature and density. Consequently, the nuclear fusion power will drop. This has a significant impact on the performance of future nuclear fusion power plants. In this work, behaviors of sawtooth oscillations during an electron-cyclotron resonant heating (ECRH) in HL-2A tokamak experiment are studied. The simulation of plasma in HL-2A tokamak is carried out using the 1.5 D BALDUR integrated predictive modeling code, where the plasma core can be described by the combination of anomalous and neoclassical transport. This simulation used the Mixed Bohm/Gyro-Bohm (Mixed B/gB) model for the anomalous transport and the the NCLASS module for the neoclassical transport. For the anomamouse transport, we use Multimode (MMM95) model, while for the neoclassical transport, we use the NCLASS module for the neoclassical transport. In each simulation, a sawtooth crash is predicted by either Rogers-Zakharov sawtooth triggering model, Park-Monticello sawtooth triggering model, or Porcelli sawtooth triggering model. The effect of sawtooth crash is described by a modified Kadomtsev magnetic reconnection model

  10. Scaling law of runaway electrons in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Zheng Yongzhen

    2005-01-01

    Runaway confinement time in ohmic and additionally heated tokamak plasmas presents an anomalous behavior in comparison with theoretical predictions based on neoclassical models. A one-dimensional numerical including generation and loss effects for runaway electrons is used to deduce the dependence of runaway energy ε τ on runaway confinement time. The simulation results are presented in the form of a scaling law for ε τ on plasma parameters. The scaling of ε τ and therefore the runaway confinement time and runaway electron diffusivity has been studied in the HL-1M tokamak, by measuring hard X-ray spectra under different experimental conditions. (authors)

  11. Effect of recycling in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Zheng Yongzhen

    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 x 10 20 m -3 is reached in HL-1M. (author)

  12. Pulse discharge cleaning of the vacuum vessel of HL-1 tokamak

    International Nuclear Information System (INIS)

    Li Guodong; Zhu Yukun; Xiao Zhenggui; Sun Shouqi; Ze Mingrui

    1986-01-01

    The HL-1 Tokamak was test-operated on September 21, 1984. During the period of vacuum conditioning, including 60 hours of baking up to 200 deg C and 7 x 10 4 shots of pulse discharge cleaning, the calculated quantities of carbon and oxygen removed are equivalent to 24 and 6 monolayers, respectively. Then, 124 shots of tokamak discharge were performed with low level plasma parameters. The plasma current and pulse length achieved were 60 kA and 85 ms at the toroidal magnetic field of 15 kG. This paper described the techniques used and the effect on discharge characteristics of bakeout and pulse discharge cleaning of the vacuum vessel

  13. Acceleration optimization of real-time equilibrium reconstruction for HL-2A tokamak discharge control

    Science.gov (United States)

    Rui, MA; Fan, XIA; Fei, LING; Jiaxian, LI

    2018-02-01

    Real-time equilibrium reconstruction is crucially important for plasma shape control in the process of tokamak plasma discharge. However, as the reconstruction algorithm is computationally intensive, it is very difficult to improve its accuracy and reduce the computation time, and some optimizations need to be done. This article describes the three most important aspects of this optimization: (1) compiler optimization; (2) some optimization for middle-scale matrix multiplication on the graphic processing unit and an algorithm which can solve the block tri-diagonal linear system efficiently in parallel; (3) a new algorithm to locate the X&O point on the central processing unit. A static test proves the correctness and a dynamic test proves the feasibility of using the new code for real-time reconstruction with 129 × 129 grids; it can complete one iteration around 575 μs for each equilibrium reconstruction. The plasma displacements from real-time equilibrium reconstruction are compared with the experimental measurements, and the calculated results are consistent with the measured ones, which can be used as a reference for the real-time control of HL-2A discharge.

  14. Overview of HL-2A experiment results

    International Nuclear Information System (INIS)

    Yang, Q.W.; Yong Liu; Ding, X.T.

    2007-01-01

    Recent experiment results from the HL-2A tokamak are presented in this paper. Supersonic molecular beam injection (SMBI) with liquid nitrogen temperature propellant is used. Low temperature SMBI can form hydrogen clusters that penetrate into the plasma more deeply and efficiently. Particle diffusion coefficient and convection velocity (D = 0.5-1.5 m 2 s -1 and V conv -1 , respectively) are obtained at the plasma periphery using modulated SMBI. Multi-probe measurements reveal the m = 0-1, n = 0 symmetries of directly measured low frequency (7-9 kHz) electric potential and field are simultaneously observed for the first time. Impurity transport is determined with the laser blow-off system and transport code. A disruption predictor has been derived based on MHD activity observations and statistical analysis. Sawtooth characteristics during ECRH are investigated and coupling between m = 1 and m/n = 2/1 modes is studied. Detachment features of HL-2A divertor are numerically and experimentally studied using the code SOLPS5.0 and measured data. The long divertor legs and thin divertor throats in HL-2A pose MHD shaping problems resulting in momentum losses even at low densities and strongly enhanced main chamber losses

  15. Observation of electron temperature profile in HL-1M tokamak

    International Nuclear Information System (INIS)

    Cao Jianyong; Xu Deming; Ding Xuantong

    2000-01-01

    The principle and method of the electron temperature measurement by means of electron cyclotron emission (ECE) have been described. Several results under different conditions on HL-1M tokamak have been given. The hollow profile of electron temperature appears in some stages, such as current rising, pellet injection and impurity concentration in the plasma centre. When the bias voltage is applied, the electron temperature profile become steeper. All of the phenomena are related with the transport in plasma centre

  16. Low–intermediate–high confinement transition in HL-2A tokamak plasmas

    International Nuclear Information System (INIS)

    Cheng, J.; Dong, J.Q.; Yan, L.W.; Hong, W.Y.; Zhao, K.J.; Huang, Z.H.; Ji, X.Q.; Zhong, W.L.; Yu, D.L.; Nie, L.; Song, X.M.; Yang, Q.W.; Ding, X.T.; Duan, X.R.; Liu, Yong; Itoh, K.; Itoh, S.-I.; Zou, X.L.

    2014-01-01

    The dynamics of low–intermediate–high confinement transitions was studied using a four-step Langmuir probe in the HL-2A edge plasma. Two types (dubbed type-Y and type-J) of limit cycle oscillations (LCOs) with opposite temporal ordering between the radial electric field and turbulence were first observed. In type-Y, the turbulence grows first, followed by the localized electric field. In contrast, the electric field leads turbulence in type-J. In addition, the Reynolds stress gradient is found not enough to drive the LCO flow and the three-wave nonlinear coupling is weak there. The continuously increasing amplitude of magnetic fluctuations and the significant correlation between the magnetic fluctuation and the electron pressure gradient indicate an important role of diamagnetic drifts in the L–H transition. Mode numbers of magnetic fluctuations in the LCO frequency are identified to be m/n = 1/0. (paper)

  17. Simulation of Plasma Disruptions for HL-2M with the DINA Code

    International Nuclear Information System (INIS)

    Xue Lei; Duan Xu-Ru; Zheng Guo-Yao; Yan Shi-Lei; Liu Yue-Qiang; Dokuka, V. V.; Khayrutdinov, R. R.; Lukash, V. E.

    2015-01-01

    Plasma disruption is often an unavoidable aspect of tokamak operations. It may cause severe damage to in-vessel components such as the vacuum vessel conductors, the first wall and the divertor target plates. Two types of disruption, the hot-plasma vertical displacement event and the major disruption with a cold-plasma vertical displacement event, are simulated by the DINA code for HL-2M. The time evolutions of the plasma current, the halo current, the magnetic axis, the minor radius, the elongation as well as the electromagnetic force and eddy currents on the vacuum vessel during the thermal quench and the current quench are investigated. By comparing the electromagnetic forces before and after the disruption, we find that the disruption causes great damage to the vacuum vessel conductors. In addition, the hot-plasma vertical displacement event is more dangerous than the major disruption with the cold-plasma vertical displacement event. (paper)

  18. Three-dimensional features of GAM zonal flows in the HL-2A tokamak

    International Nuclear Information System (INIS)

    Yan, L.W.; Cheng, J.; Hong, W.Y.; Zhao, K.J.; Lan, T.; Dong, J.Q.; Liu, A.D.; Yu, C.X.; Yu, D.L.; Qian, J.; Huang, Y.; Yang, Q.W.; Ding, X.T.; Liu, Y.; Pan, C.H.

    2007-01-01

    A novel design of the three-step Langmuir probe (TSLP) array has been developed to investigate the zonal flow (ZF) physics in the HL-2A tokamak. Three TSLP arrays are applied to measure the three-dimensional (3D) features of ZFs. They are separated by 65 mm in the poloidal and 800 mm in the toroidal directions, respectively. The 3D properties of the geodesic acoustic mode (GAM) ZFs are presented. The poloidal and toroidal modes of the radial electric fields of the GAM perturbations are simultaneously determined in the HL-2A tokamak for the first time. The modes have narrow radial wave numbers (k r ρ i = 0.03-0.07) and short radial scale lengths (2.4-4.2 cm). High coherence of both the GAM and the ambient turbulence separated by toroidal 22.5 0 along a magnetic field line is observed, which contrasts with the high coherence of the GAM and the low coherence of the ambient turbulence apart from the field line. The nonlinear three wave coupling between the turbulent fluctuations and the ZFs is a plausible mechanism for flow generation. The skewness and kurtosis spectra of the probability distribution function of the potential perturbations are contrasted with the corresponding bicoherence for the first time, which support the three wave coupling mechanism

  19. Development of a USB-based multi-channel time division scaler for HL-2A

    International Nuclear Information System (INIS)

    Liang Ping

    2008-01-01

    HL-2A is China's first Tokamak device with divertor configuration. Mastering the process and plasma parameter changes with time are of great significance to achieve controlled nuclear fusion. In the recent upgrading of HL-2A, for which a higher and faster electronic equipment was required, we developed a new type USB multi-channel time division scaler for HL-2A including functions: USB interface, PC graphical user interface, simultaneously calibrating more than five channel signals, optional time division spacing 2-50 ms, count rate up to over 2 MHz, accessing all the information 20 s after a activated signal, and processing data and displaying off-line. (authors)

  20. Optics of ion beams for the neutral beam injection system on HL-2A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zou, G. Q.; Lei, G. J.; Cao, J. Y.; Duan, X. R. [Southwestern Institute of Physics, Chengdu, 610041 (China)

    2012-07-15

    The ion beam optics for the neutral beam injection system on HL-2A Tokomak is studied by two- dimensional numerical simulation program firstly, where the emitting surface is taken at 100 Debye lengths from the plasma electrode. The mathematical formulation, computation techniques are described. Typical ion orbits, equipotential contours, and emittance diagram are shown. For a fixed geometry electrode, the effect of plasma density, plasma potential and plasma electron temperature on ion beam optics is examined, and the calculation reliability is confirmed by experimental results. In order to improve ion beam optics, the application of a small pre-acceleration voltage ({approx}100 V) between the plasma electrode and the arc discharge anode is reasonable, and a lower plasma electron temperature is desired. The results allow optimization of the ion beam optics in the neutral beam injection system on HL-2A Tokomak and provide guidelines for designing future neutral beam injection system on HL-2M Tokomak.

  1. Optics of ion beams for the neutral beam injection system on HL-2A Tokamak.

    Science.gov (United States)

    Zou, G Q; Lei, G J; Cao, J Y; Duan, X R

    2012-07-01

    The ion beam optics for the neutral beam injection system on HL-2A Tokomak is studied by two- dimensional numerical simulation program firstly, where the emitting surface is taken at 100 Debye lengths from the plasma electrode. The mathematical formulation, computation techniques are described. Typical ion orbits, equipotential contours, and emittance diagram are shown. For a fixed geometry electrode, the effect of plasma density, plasma potential and plasma electron temperature on ion beam optics is examined, and the calculation reliability is confirmed by experimental results. In order to improve ion beam optics, the application of a small pre-acceleration voltage (∼100 V) between the plasma electrode and the arc discharge anode is reasonable, and a lower plasma electron temperature is desired. The results allow optimization of the ion beam optics in the neutral beam injection system on HL-2A Tokomak and provide guidelines for designing future neutral beam injection system on HL-2M Tokomak.

  2. Profiles of zonal flows and turbulence mode numbers and probe system in the HL-2A tokamak

    International Nuclear Information System (INIS)

    Hong Wenyu; Zhao Kaijun; Yan Longwen; Dong Jiaqi; Cheng Jun; Qian Jun

    2009-01-01

    The toroidal and poloidal symmetries (m-0, n-0) of the measured low frequency zonal flows (f=0-5 kHz) and geodesic acoustic mode zonal flow (f=16 kHz) electric potential and radial promulgate features were unambiguously identified with displaced Langmuir probe arrays in the edge plasma of the HL-2A tokamak for the first time. The finite radial wave vector (K r-LF =0.6 cm -1 , K r-GAM =2 cm -1 ) of the flows was simultaneously estimated. The formation mechanism of the flows is identified to be nonlinear three wave coupling between high frequency turbulent fluctuations and the flows. Changes of zonal flow amplitude bring by ECRH power and the boundary safety factors were simply studied. Moreover, change of zonal flow amplitude in radial direction was too observed. (authors)

  3. A new high sensitivity far-infrared laser interferometer for the HL-2A tokamak

    Science.gov (United States)

    Li, Y. G.; Zhou, Y.; Li, Y.; Deng, Z. C.; Wang, H. X.; Yi, J.

    2017-08-01

    A new four-chord Michelson-type formic acid (HCOOH, λ = 432.5 μm) laser interferometer has been successfully commissioned on the HL-2A tokamak to measure the electron density and density fluctuations. Due to the employment of the two-laser heterodyne technique, the time resolution of the interferometer reached 1.0 microseconds (μs). Four chords of line electron densities with a line-averaged density resolution 2 × 1016/m3 were obtained in a recent HL-2A experimental campaign, and detailed electron density fluctuations, caused by events such as edge localized mode, sawtooth precursor-oscillations, and energetic particle driven instabilities, were distinctly measured. In particular, the high-frequency electron density fluctuations (up to 500 kHz) caused by the reversed shear Alfvénic eigenmode were observed by the internal two interferometry channels, and their fluctuation location could be approximately identified from the spectra characteristics of multi-chord line electron densities.

  4. Density and impurity profile behaviours in HL-2A tokamak with different gas fuelling methods

    International Nuclear Information System (INIS)

    Zheng-Ying, Cui; Yan, Zhou; Wei, Li; Bei-Bin, Feng; Ping, Sun; Chun-Feng, Dong; Yi, Liu; Wen-Yu, Hong; Qing-Wei, Yang; Xuan-Tong, Ding; Xu-Ru, Duan

    2009-01-01

    The electron density profile peaking and the impurity accumulation in the HL-2A tokamak plasma are observed when three kinds of fuelling methods are separately used at different fuelling particle locations. The density profile becomes more peaked when the line-averaged electron density approaches the Greenwald density limit n G and, consequently, impurity accumulation is often observed. A linear increase regime in the density range n e G and a saturation regime in n e > 0.6n G are obtained. There is no significant difference in achieved density peaking factor f ne between the supersonic molecular beam injection (SMBI) and gas puffing into the plasma main chamber. However, the achieved f ne is relatively low, in particular, in the case of density below 0.7n G , when the working gas is puffed into the divertor chamber. A discharge with a density as high as 1.2n G , i.e. n e = 1.2n G , can be achieved by SMBI just after siliconization as a wall conditioning. The metallic impurities, such as iron and chromium, also increase remarkably when the impurity accumulation happens. The mechanism behind the density peaking and impurity accumulation is studied by investigating both the density peaking factor versus the effective collisionality and the radiation peaking versus density peaking. (fluids, plasmas and electric discharges)

  5. Effects of fuelling by using high-pressure supersonic molecular beam in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Feng Beibin; Feng Zhen; Dong Jiafu; Li Wenzhong; Xu Deming; Hong Wenyu

    2002-01-01

    Supersonic molecular beam (SMB), as a new fuelling method, has been successfully developed and used in HL-1M tokamak and HT-7 superconducting tokamak. The hydrogen clusters have been found in the beam produced by high working-gas pressure in recent experiments. With a penetration depth of hydrogen particles greater than 17 cm, the rate of increase of electron density for SMB injection, dn e -bar/dt, approaches that of the small ice pellet injection. The plasma density increases step by step after multi-pulse SMB injection, just as multi-pellet fuelling. Comparison of fuelling effects was made between SMB and ice pellet injection on the same shot of ohmic discharge in HL-1M

  6. The ohmic heating power supply for HL-1 tokamak

    International Nuclear Information System (INIS)

    Mingrui, Z.; Jiashun, C.

    1986-01-01

    A combination of capacitor banks, inductor and DC Fly wheel-Generator sets are used as ohmic heating power supply (OHPS) for HL-1, which is the largest tokamak in China. This system can give changeable waveform of current in a simple way, because of the use of protection for capacitor banks by changeable connection in easy way. Since the technology of forced zero current in the commutating breaker and synchronous self-triggering crowbar are used, the smooth conversion between the wave front provided by discharge of the capacitor banks and the flat top sustained by the inductor and flywheel realized. The performance of the system was tested by a dummy load and the system has been used in the HL-1 experiments. It is confirmed that this system is sufficiently available for the ohmic heating and has important effects on the long plasma lasting time on the order of 1 sec

  7. Predictive study on high performance modes of operation in HL-2A

    International Nuclear Information System (INIS)

    Gao Qingdi; Li Fangzhu; Zhang Jinhua; Qu Hongpeng; Budny, R.V.

    2003-01-01

    High performance scenarios in the HL-2A tokamak are studied by numerical modeling. Through shifting the plasma column outwards a shaping plasma with significant triangularity is achieved with sufficient room left for the RF antenna. For the out-shifted shaping plasma, ripple loss of high energy ions during NBI is analyzed. The results show that the ripple loss fraction of NBI power for the shaping plasma is not higher than that for the unshifted circular plasma. The time dependent TRANSP code is used to model realistic reversed magnetic shear operation in HL-2A. In order to sustain the RS operation towards steady-state, off-axis current drive with a lower hybrid wave at 2.45GHz is used to control the current profile. A steady-state RS discharge is formed and sustained until the LH power is turned off; the plasma confinement is enhanced with the development of an internal transport barrier. In the RS discharges with shaping plasma geometry a double transport barrier is produced. (author)

  8. Characterization of plasma current quench during disruptions at HL-2A

    Science.gov (United States)

    Zhu, Jinxia; Zhang, Yipo; Dong, Yunbo; HL-2A Team

    2017-05-01

    The most essential assumptions of physics for the evaluation of electromagnetic forces on the plasma-facing components due to a disruption-induced eddy current are characteristics of plasma current quenches including the current quench rate or its waveforms. The characteristics of plasma current quenches at HL-2A have been analyzed during spontaneous disruptions. Both linear decay and exponential decay are found in the disruptions with the fastest current quenches. However, there are two stages of current quench in the slow current quench case. The first stage with an exponential decay and the second stage followed by a rapid linear decay. The faster current quench rate corresponds to the faster movement of plasma displacement. The parameter regimes on the current quench time and the current quench rates have been obtained from disruption statistics at HL-2A. There exists no remarkable difference for distributions obtained between the limiter and the divertor configuration. This data from HL-2A provides basic data of the derivation of design criteria for a large-sized machine during the current decay phase of the disruptions.

  9. Design and construction of high-frequency magnetic probe system on the HL-2A tokamak

    Science.gov (United States)

    Liang, S. Y.; Ji, X. Q.; Sun, T. F.; Xu, Yuan; Lu, J.; Yuan, B. S.; Ren, L. L.; Yang, Q. W.

    2017-12-01

    A high-frequency magnetic probe system is designed, calibrated and constructed on the HL-2A tokamak. To investigate the factors which affect the probe frequency response, the inductance and capacitance in the probe system are analyzed using an equivalent circuit. Suitable sizes and turn number of the coil, and the length of transmission cable are optimized based on the theory and detailed test in the calibration. To deal with the frequency response limitation and bake-out, the ceramic grooved technique is used and the probe is wound with a bare copper wire. A cascade filter is manufactured with a suitable bandwidth as well as a good phase consistency between channels. The system has been used in the experiment to measure high frequency (≤300 kHz) magnetohydrodynamic fluctuations, which can meet the requirement of physical analysis on HL-2A.

  10. The preliminary studies of HL-1 plasma with pump limiter

    International Nuclear Information System (INIS)

    Ran, L.B.; Li, G.D.; Sum, S.Q.

    1991-01-01

    Reduction of the impurity and control of the density are very important task in the present tokamak research. Some experiments showed that the pump limiter is effective to improve impurity and density control. A small-scale pump limiter has been installed on HL-1 tokamak in 1990 and the experiment has been carried out firstly with hydrogen ohmic discharge. In this paper we would present the improvements of the plasma properties as the pump limiter was main limiter and the movable limiter was auxiliary limiter. (author) 3 refs., 8 figs

  11. Simulation of runaway electron generation and diffusion during major disruptions in the HL-2A tokamak

    International Nuclear Information System (INIS)

    Li, Yanli; Sun, Jizhong; Zhang, Yipo; Sang, Chaofeng; Wu, Na; Wang, Dezhen

    2014-01-01

    Highlights: • The strong and long duration magnetic perturbation (δB/B ∼ 1.0 × 10 −3 ) can restrain the RE generation effectively. • The REs are generated initially in the plasma core during disruptions. • The toroidal electric field does not exhibit a centrally hollow phenomenon. • The toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field. - Abstract: The generation and diffusion of runaway electrons (REs) during major disruptions in the HL-2A tokamak has been studied numerically. The diffusion caused by the magnetic perturbation is especially addressed. The simulation results show that the strong magnetic perturbation (δB/B ∼ 1.0 × 10 −3 ) can cause a significant loss of REs due to the radial diffusion and restrain the RE avalanche effectively. The results also indicate that the REs are generated initially in the plasma core during disruptions, and that the toroidal electric field does not exhibit a centrally hollow phenomenon. In addition, it is found that the toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field

  12. Neutral pressure measured by fast ionization gauge in HL-2A

    International Nuclear Information System (INIS)

    Wang, M.X.; Yan, L.W.; Hong, W.Y.; Cui, Z.Y.; Pan, Y.D.; Yao, L.H.; Cao, Z.; Yang, Q.W.; Liu, Y.

    2007-01-01

    A fast ionization gauge capable of working in a strong magnetic field and noisy environment is developed for the pressure measurement in the main chamber and the divertor chamber of the HL-2A tokamak. The neutral pressure in the main chamber evidently decreases after in-situ siliconization. Improvement of boundary plasma confinement is observed with SMBI or PI fueling after siliconization. The neutral pressure in the main chamber for divertor configuration is 50% lower than that for limiter one. The pressure in the divertor chamber rises when the strike point moves toward the bottom of the vertical target plates. The maximum pressure compression of is 24. The relative high pressure in the main chamber is contributed primarily by plasma limiter interaction due to the short distance from separatrix to limiter and secondly by neutral flux from the divertor chamber to the main chamber by leaks in HL-2A

  13. Preliminary study of energy confinement data with a statistical analysis system in HL-2A tokamak

    International Nuclear Information System (INIS)

    Xu Yuan; Cui Zhengying; Ji Xiaoquan; Dong Chunfeng; Yang Qingwei; O J W F Kardaun

    2010-01-01

    Taking advantage of the HL-2A experimental data,an energy confinement database facing ITERL DB2.0 version has been originally established. As for this database,a world widely used statistical analysis system (SAS) has been adopted for the first time to analyze and evaluate the confinement data from HL-2A and the research on scaling laws of energy confinement time corresponding to plasma density is developed, some preliminary results having been achieved. Finally, through comparing with both ITER scaling law and previous ASDEX database, the investigation about L-mode confinement quality on HL-2A and influence of temperature on Spitzer resistivity will be discussed. (authors)

  14. A new way of soft X-ray spectrum measurement in the HL-2A tokamak

    International Nuclear Information System (INIS)

    Liao Min; Chen Liaoyuan; Song Xianying; Li Yongge; Luo Cuiwen; Pan Wei; Pan Li

    2007-01-01

    A soft X-ray spectrum measurement system is newly developed for the real-time acquisition, data processing and display of electron temperature. In this system, a personal computer with high-speed A/D acquisition card is used to obtain the original analog data and processing software is specially developed to calculate the electron temperature. Compared with PHA (Pulse Height Analyzer), the new system provides a way to set channel number and time solution by software. This method has been applied in the HL-2A tokamak experiment, and the result shows a good agreement with the electron temperature tested by other measurements. (authors)

  15. First observation of neoclassical tearing modes in the HL-2A tokamak

    International Nuclear Information System (INIS)

    Ji Xiaoquan; Yang Qingwei; Liu Yi; Zhou Jun; Feng Beibin; Yuan Baoshan

    2010-01-01

    The m/n=2/1 neoclassical tearing modes are first observed in HL-2A low density, low beta plasma with central electron cyclotron resonance heating. The neoclassical tearing modes (NTMs) are triggered by a sawtooth crash with m/n=1/1 precursors, which are toroidal coupled with a small scale m/n=2/1 mode. The time history of the island width is compared with the prediction of the NTM theory, showing a good agreement between theory and experiment. (authors)

  16. Modelling of profile control with LH wave injection in the HL-2A single-null divertor plasma

    International Nuclear Information System (INIS)

    Gao Qingdi; Yuan Baoshan; Li Fangzhu; Wang, Aike; Budny, R.V.

    2005-01-01

    In the HL-2A tokamak a single-null divertor configuration has been established. The separatrix of the single-null diverted plasma was identified with a filament model, and the determined striking area on the target plate is in agreement with the measurements of electric probe array. Higher power LH wave (1.5MW) is injected to the diverted plasma with a nearly symmetric spectrum. Dominant electron heating and current profile control are investigated with numerical simulation. Plasma heating by electron Landau interaction results in operation scenarios of preferentially dominant electron heating. Due to the off-axis driven current, an optimized q-profile is formed, and an enhanced confinement regime with steep electron temperature gradient is produced. The clear decrease of the electron thermal conductivity in the LH power deposition region shows that an electron-ITB is developed. When higher LH power injects into the target plasma that is heated by NBI (0.5MW), the ion temperature has a large increment in addition to the high increase of electron temperature. The temperature profiles indicate that an enhanced core confinement is established with both ion-ITB and electron-ITB developed. (author)

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

    Science.gov (United States)

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

    2015-02-01

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

  18. HL-1 tokamak data acquisition system and its initial application in the physical experiment

    International Nuclear Information System (INIS)

    Deng Huichen; Fu Bo; Dong Jiafu

    1989-11-01

    A HL-1 tokamak data acquisition system has been developed and has been used in the physical experiment. The hardware and software configuration of the system, as well as the typical acquired data in the HL-1 experiment are introduced

  19. Experiment of laser thomson scattering at HL-1 tokamak device

    International Nuclear Information System (INIS)

    Zuo Henian; Chen Jiafu; Yan Derong; Liu Aiping; Shi Peilan; Wang Wei; Liu Xiaomei

    1989-05-01

    The structure and performance of the Ruby Laser Thomson Scattering apparatus for HL-1 tokamak device is described. The method of acquisition and calibration of multichannel scattered signals are presented. Examples of measured electron temperature T. with experimental error are given

  20. High performance experiments on high pressure supersonic molecular beam injection in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Dong Jiafu; Zhou Yan; Feng Beibing; Cao Jianyong; Li Wei; Feng Zhen; Zhang Jiquan; Hong Wenyu; Cui Zhengying; Wang Enyao; Liu Yong

    2004-01-01

    Supersonic molecular beam injection (SMBI) was first proposed and demonstrated on the HL-1 tokamak and was successfully developed and used on HL-1M. Recently, new results of SMBI experiments were obtained by increasing the gas pressure from 0.5 to over 1.0 MPa. A stair-shaped density increment was obtained with high-pressure multi-pulse SMBI that was similar to the density evolution behaviour during multi-pellet injection. This demonstrated the effectiveness of SMBI as a promising fuelling tool for steady-state operation. The penetration depth and injection speed of the high-pressure SMBI were roughly measured from the contour plot of the Hα emission intensity. It was shown that injected particles could penetrate into the core region of the plasma. The penetration speed of high-pressure SMBI particles in the plasma was estimated to be about 1200 m s -1 . In addition, clusters within the beam may play an important role in the deeper injection. (author)

  1. Multi-channel time-division integrator in HL-2A

    International Nuclear Information System (INIS)

    Yan Ji

    2008-01-01

    HL-2A is China's first Tokamak device with divertor configuration (magnetic confinement controlled nuclear fusion device). To find out the details of on-going fusion reaction at different times is of important significance in achieving controlled nuclear fusion. We developed a new type multi-channel time-division integrator for HL-2A. It has functions of automatic cutting off negative pulse of the input signals, optional integrating time division spacing 0.2-1 ms, TTL starting trigger signal, automatic regularly work 20 s, and integrating 10 channel at the same time. (authors)

  2. Density Limits with Different Fuelling Methods in the HL-2A Tokamak

    International Nuclear Information System (INIS)

    Yan Longwen; Zhou Yan; Chen Chengyuan; Cao Zeng; Song Xianming; Li Wei; Dong Yunbo; Hong Wenyu; Yang Qingwei; Duan Xuru

    2009-01-01

    Density limits with different fuelling methods have been compared in HL-2A, i.e. direct gas puffing and supersonic molecular beam injection (SMBI) from outer midplane, and divertor gas fuelling. The maximum densities for low current discharges are 3.4x10 19 m -3 , 4.3x10 19 m -3 and 4.7x10 19 m -3 for the 3 kinds of fuelling methods. The corresponding density ratios to Greenwald density limit are 0.9, 1.1, 1.2, respectively. The behavior of density limit disruption is analyzed as well. (magnetically confined plasma)

  3. H-L transition caused by irregular probe insertion in the JFT-2M tokamak

    International Nuclear Information System (INIS)

    Hanada, K.; Hasegawa, M.; Shinohara, K.

    1996-01-01

    A pair of probe pins whose lengths differ by 5 mm are inserted into the separatrix of H-mode plasmas. When the long pin is biased positively against the short one, the H-L transition takes place and the current through the resistor making a connection between these pins increases for 10 μs and the potential of the pin increases for 100 μs. The energy dissipated by the resistor during the H-L transition is comparable to the energy of the radial electric field in H-mode plasma. When the long pin is biased negatively, the H-L transition also occurs; however, it takes 1 ms to change the potential of the pin and the plasma returns to H-mode after the extraction of the probe. (author)

  4. Study of heat transfer and particle transport in Tore Supra and HL-2A tokamaks

    International Nuclear Information System (INIS)

    Song, S.

    2011-12-01

    This thesis reports on experimental studies of heat and particles transport performed on 2 large tokamaks: Tore Supra (based at CEA/Cadarache, France) and HL-2A (based at the Southwestern Institute of Physics, Chengdu, China). The modulated source is the Electron Cyclotron Resonance Heating (ECRH) for the heat pinch and density pump-out studies, while the non-local transport experiments use the Supersonic Molecular Beam Injection (SMBI) as source of modulation. The emphasis is put on the inward heat pinch. In the off-axis ECRH modulation experiments on Tore Supra with low frequency (1 Hz), strong heat inward transport has been observed, in particular for low density. Two transport models have been applied in order to analyze the experimental behavior. The first one is the linear pinch model (LPM) and the second one is an empirical model based on micro-instabilities theory, named Critical Gradient Model (CGM). Good agreement has been found for all harmonics between the experimental data and the simulation using LPM. On the other hand, good agreement has not been achieved using CGM. The density pump-out with large particles and energy losses during ECRH is commonly observed in tokamaks. A new dynamic approach using the modulation technique has been used in HL-2A for analyzing the transient phase of the density pump-out. A correlation between the turbulence increase and the density pump-out has been found. The non-local transport phenomenon, characterized by a fast transient process compared to the normal diffusive response to the perturbation is observed. Both phenomena, i.e., pump-out and non-locality, show as simultaneous variation of density and temperature. This can be an inspiration for the usage of a transport matrix which considers the density and temperature evolution together. Simulations with a simple transport matrix, with non-diagonal terms coupling temperature and density qualitatively reproduce the non-local and pump-out effects qualitatively

  5. Study of the Polarization Strategy for Electron Cyclotron Heating Systems on HL-2M

    Science.gov (United States)

    Zhang, F.; Huang, M.; Xia, D. H.; Song, S. D.; Wang, J. Q.; Huang, B.; Wang, H.

    2016-06-01

    As important components integrated in transmission lines of electron cyclotron heating systems, polarizers are mainly used to obtain the desired polarization for highly efficient coupling between electron cyclotron waves and plasma. The polarization strategy for 105-GHz electron cyclotron heating systems of HL-2M tokamak is studied in this paper. Considering the polarizers need high efficiency, stability, and low loss to realize any polarization states, two sinusoidal-grooved polarizers, which include a linear polarizer and an elliptical polarizer, are designed with the coordinate transformation method. The parameters, the period p and the depth d, of two sinusoidal-grooved polarizers are optimized by a phase difference analysis method to achieve an almost arbitrary polarization. Finally, the optimized polarizers are manufactured and their polarization characteristics are tested with a low-power test platform. The experimental results agree well with the numerical calculations, indicating that the designed polarizers can meet the polarization requirements of the electron cyclotron heating systems of HL-2M tokamak.

  6. Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

    Science.gov (United States)

    Zhang, Y. P.; Liu, Yi; Luo, X. B.; Isobe, M.; Yuan, G. L.; Liu, Y. Q.; Hua, Y.; Song, X. Y.; Yang, J. W.; Li, X.; Chen, W.; Li, Y.; Yan, L. W.; Song, X. M.; Yang, Q. W.; Duan, X. R.

    2014-05-01

    A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y. Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team, Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported.

  7. Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Y. P., E-mail: zhangyp@swip.ac.cn; Liu, Yi; Yuan, G. L.; Song, X. Y.; Yang, J. W.; Li, X.; Chen, W.; Li, Y.; Yan, L. W.; Song, X. M.; Yang, Q. W.; Duan, X. R. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Luo, X. B.; Liu, Y. Q.; Hua, Y. [Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610041 (China); Isobe, M. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5259 (Japan)

    2014-05-15

    A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y. Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team , Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported.

  8. Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

    International Nuclear Information System (INIS)

    Zhang, Y. P.; Liu, Yi; Yuan, G. L.; Song, X. Y.; Yang, J. W.; Li, X.; Chen, W.; Li, Y.; Yan, L. W.; Song, X. M.; Yang, Q. W.; Duan, X. R.; Luo, X. B.; Liu, Y. Q.; Hua, Y.; Isobe, M.

    2014-01-01

    A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y. Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team , Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported

  9. The Supervisory Control System for the HL-2A Neutral Beam Injector

    Science.gov (United States)

    Li, Bo; Li, Li; Feng, Kun; Wang, Xueyun; Yang, Jiaxing; Huang, Zhihui; Kang, Zihua; Wang, Mingwei; Zhang, Guoqing; Lei, Guangjiu; Rao, Jun

    2009-06-01

    Supervisory control and protection system of the neutral beam injector (NBI) in the HL-2A tokamak is presented. The system is used for a safe coordination of all the main NBI subsystems. Because the system is based on computer networks with its transmission medium of optical fiber, its advantages in high operational stability, reliability, security and flexible functional expandability are clearly shown during the NBI commissioning and heating experiment in HL-2A.

  10. Physics design and experimental study of tokamak divertor

    International Nuclear Information System (INIS)

    Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

    2007-06-01

    The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

  11. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Furth, H.P.

    1985-10-01

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

  12. Research on the phase adjustment method for dispersion interferometer on HL-2A tokamak

    Science.gov (United States)

    Tongyu, WU; Wei, ZHANG; Haoxi, WANG; Yan, ZHOU; Zejie, YIN

    2018-06-01

    A synchronous demodulation system is proposed and deployed for CO2 dispersion interferometer on HL-2A, which aims at high plasma density measurements and real-time feedback control. In order to make sure that the demodulator and the interferometer signal are synchronous in phase, a phase adjustment (PA) method has been developed for the demodulation system. The method takes advantages of the field programmable gate array parallel and pipeline process capabilities to carry out high performance and low latency PA. Some experimental results presented show that the PA method is crucial to the synchronous demodulation system and reliable to follow the fast change of the electron density. The system can measure the line-integrated density with a high precision of 2.0 × 1018 m‑2.

  13. HL-2A experiment and ITER-related activity at SWIP

    International Nuclear Information System (INIS)

    Duan Xuru

    2007-01-01

    In this overview the recent progress on HL-2A tokamak experiment and ITER-related activity at SWIP is summarized. Experiment on HL-2A is one of the important research activities at SWIP. In the last two years, some new hardware had been developed, these include four sets of ECRH system with a total power up to 2 MW, new diagnostics such as 8-channel laser interferometer. The studied subjects were focused on plasma auxiliary heating, fuelling, transport, edge plasma physics and turbulence, etc. Progress in these fields has been obtained. For example, the toroidal symmetry of the geodesic acoustic mode (GAM), the oscillating branch of zonal flows has been demonstrated for the first time using a novel 3-step Langmuir Probe, and the poloidal and radial structure of the low frequency electric potential and field were simultaneously observed. During ECRH experiments under different discharge conditions, the MHD instability excited by high energetic electrons was investigated. Besides, non-local heat transport due to SMBI during ECRH was studied. Another important fusion activity at SWIP is the ITER relevant technology. The R and D of four ITER procurements (first wall and shielding blanket, magnet gravity support, gas injection and glow discharge cleaning system, neutron flux measurement) has been undertaken. Progress has been made, e.g. the technology for manufacturing high purity (>99%) ITER specified Be plate and CuCrZr alloy is obtained, their major mechanical and physical properties were measured. For ITER-TBM, a structural material named as CLF-1, a type of reduced activation ferritic/martenstic steel, was developed. Besides, some progress in fusion reactor design and related technology was achieved. (authors)

  14. Time-resolved Evolution of Low Frequency Electrostatic Fluctuations during Slow L-H Transition at the Boundary Plasma of HL-2A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, J.; Zhao, K. J.; Li, Y. G.; Song, X. M.; Yang, Q. W.; Ding, X. T.; Duan, X. R.; Liu, Y., E-mail: chengj@swip.ac.cn [Southwestern Institute of Physics, Chengdu (China); Huang, Z. H.; Yan, L. W.; Dong, J. Q.; Hong, W. Y.; Kong, D. F.; Lan, T.; Liu, A. D. [Southwestern Institute of Physics, Hefei (China); Xu, M. [CMTFO and CER and MAE Department, UCSD, La Jolla (United States)

    2012-09-15

    Full text: A quasi-period electrostatic oscillation with 2 - 3 kHz is observed using Langmuir probe array during slow L-H transition in edge plasma of HL-2A Tokomak. This low frequency oscillation radially propagates inwards with 0.3 - 0.6 km/s inside the separatrix about 3 - 8 mm, and it appears on potential, density, electron pressure and Reynolds stress gradients. The dP{sub e}/dr fluctuation amplitude can reach 30 - 40%. The dR{sub s}/dr is prior to E{sub r} fluctuation about {pi}/2, indicating the existence of nonlinear interaction between them. In near SOL, this low frequency oscillation also appears in potentials, E{sub r} and density fluctuation, suggesting a significant correlation among them at edge and near SOL plasma. This quasi-period oscillation might be correlated with mean flow or low frequency zonal flow, and the latter might set a condition for the former developing, implying a competitive process between them. The competition characterized by a quasi-period oscillation seems to determine the L-H transition. (author)

  15. Data acquisition and processing based on the multi-microcomputers data communication system for the physics experiments of HL-1 tokamak

    International Nuclear Information System (INIS)

    Ren Libo; Liu Aiping; Xu Wenbin; Ao Xueling; Gao Lanbin; Yan Derong

    1989-08-01

    The data acquisition and processing system and automatic monitor system based on the data communication of multimicrocomputers for the physics experiments of HL-1 tokamak are described. The data communications between microcomputers are completed with parallel and serial interfaces. The signals from Rokofski coils, single loop turn, soft X-ray absorb spectrometer, microwave interfermeter, bolometer, soft X-ray sawtooth Au-Si surface barrier detectors and magnetic probes have been acquired and processed. Plasma parameters have been obtained. Specially, the electron density profiles have been obtained for the first time, and the simulation about plasma radius, horizontal displacement, electron temperature and electron line-average density has been completed

  16. Concept design for the central control system of HL-2A

    International Nuclear Information System (INIS)

    Song Xianming; Li Qiang; Jiang Chao

    2001-01-01

    The design principle and basic structure of the central control system for HL-2A Tokamak are introduced. Having been limited by manpower and money, the central control system should not be too expensive and too advanced. On the other hand, because of the complexity of the machine and the difficulty the author will encounter when operating the machine, the central control system should be advanced enough. If use the same technology for HL-1M to control HL-2A, the author would fail to fulfill authors' experiment goal. The central control system consists of software and hardware. The software mainly includes: (a) system monitor and control software; (b) discharge monitor and control software; (c) network and communication software. Hardware includes: (a) PLC for machine control, personnel protection and machine protection; and (b) VME computer, for the feedback control of the discharge

  17. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1987-01-01

    It has been seen that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. This trapping of a gun-injected plasma is explained in terms of a depolarization current mechanism. A model is developed that describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/proportionalT/sup 3/2//sub e/L 2 /n/sub b/α 2 0 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions

  18. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1986-01-01

    Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

  19. Synergetic Effects of Runaway and Disruption Induced by VDE on the First Wall Damage in HL-2A

    International Nuclear Information System (INIS)

    Song Xianying; Yang Jinwei; Li Xu; Yuan Guoliang; Zhang Yipo

    2012-01-01

    The plasma facing component in HL-2A has been damaged seriously after disruption, and for this reason its operation is suspended for maintenance. The experimental phenomena and plasma configurations, calculated by the current filament code (CF-code) using the plasma parameters measured by diagnostics and the signals of the magnetic probes, confirm that the first wall is damaged by the synergetic effects of runaway electrons and disruption induced by a vertical displacement event (VDE). When the plasma column is displaced upward/downward, the strong runaway electrons normally hit the baffle plate of the MP 3 or MP 1 coil in the upper and lower divertor during the disruption, causing the baffle plates to be holed and wrinkled by the energetic runaway current, and water (for cooling or heating the baffle plates) to leak into the vacuum vessel. Another disastrous consequence is that bellows underlying the baffle plate and outside the coil of MP 3 for connecting two segments of the jacket casing pipe are punctured by arcing. The arc may be part of the halo current that forms a complete circuit. The experimental phenomena are indirect but compelling evidence for the existence of a halo current during the disruption and VDE, though the halo current has not been measured by the diagnostics in the HL-2A tokamak.

  20. Synergetic Effects of Runaway and Disruption Induced by VDE on the First Wall Damage in HL-2A

    Science.gov (United States)

    Song, Xianying; Yang, Jinwei; Li, Xu; Yuan, Guoliang; Zhang, Yipo

    2012-03-01

    The plasma facing component in HL-2A has been damaged seriously after disruption, and for this reason its operation is suspended for maintenance. The experimental phenomena and plasma configurations, calculated by the current filament code (CF-code) using the plasma parameters measured by diagnostics and the signals of the magnetic probes, confirm that the first wall is damaged by the synergetic effects of runaway electrons and disruption induced by a vertical displacement event (VDE). When the plasma column is displaced upward/downward, the strong runaway electrons normally hit the baffle plate of the MP3 or MP1 coil in the upper and lower divertor during the disruption, causing the baffle plates to be holed and wrinkled by the energetic runaway current, and water (for cooling or heating the baffle plates) to leak into the vacuum vessel. Another disastrous consequence is that bellows underlying the baffle plate and outside the coil of MP3 for connecting two segments of the jacket casing pipe are punctured by arcing. The arc may be part of the halo current that forms a complete circuit. The experimental phenomena are indirect but compelling evidence for the existence of a halo current during the disruption and VDE, though the halo current has not been measured by the diagnostics in the HL-2A tokamak.

  1. Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

    Science.gov (United States)

    Xue, L.; Duan, X. R.; Zheng, G. Y.; Liu, Y. Q.; Pan, Y. D.; Yan, S. L.; Dokuka, V. N.; Lukash, V. E.; Khayrutdinov, R. R.

    2016-05-01

    Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.

  2. Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

    International Nuclear Information System (INIS)

    Xue, L; Duan, X R; Zheng, G Y; Liu, Y Q; Pan, Y D; Yan, S L; Dokuka, V N; Khayrutdinov, R R; Lukash, V E

    2016-01-01

    Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench. (paper)

  3. Control of edge localized modes by pedestal deposited impurity in the HL-2A tokamak

    Science.gov (United States)

    Zhang, Y. P.; Mazon, D.; Zou, X. L.; Zhong, W. L.; Gao, J. M.; Zhang, K.; Sun, P.; Dong, C. F.; Cui, Z. Y.; Liu, Yi; Shi, Z. B.; Yu, D. L.; Cheng, J.; Jiang, M.; Xu, J. Q.; Isobe, M.; Xiao, G. L.; Chen, W.; Song, S. D.; Bai, X. Y.; Zhang, P. F.; Yuan, G. L.; Ji, X. Q.; Li, Y. G.; Zhou, Y.; Delpech, L.; Ekedahl, A.; Giruzzi, G.; Hoang, T.; Peysson, Y.; Song, X. M.; Song, X. Y.; Li, X.; Ding, X. T.; Dong, J. Q.; Yang, Q. W.; Xu, M.; Duan, X. R.; Liu, Y.; the HL-2A Team

    2018-04-01

    Effect of the pedestal deposited impurity on the edge-localized mode (ELM) behaviour has been observed and intensively investigated in the HL-2A tokamak. Impurities have been externally seeded by a newly developed laser blow-off (LBO) system. Both mitigation and suppression of ELMs have been realized by LBO-seeded impurity. Measurements have shown that the LBO-seeded impurity particles are mainly deposited in the pedestal region. During the ELM mitigation phase, the pedestal density fluctuation is significantly increased, indicating that the ELM mitigation may be achieved by the enhancement of the pedestal transport. The transition from ELM mitigation to ELM suppression was triggered when the number of the LBO-seeded impurity exceeds a threshold value. During the ELM suppression phase, a harmonic coherent mode (HCM) is excited by the LBO-seeded impurity, and the pedestal density fluctuation is significantly decreased, the electron density is continuously increased, implying that HCM may reduce the pedestal turbulence, suppress ELMs, increase the pedestal pressure, thus extending the Peeling-Ballooning instability limit. It has been found that the occurance of the ELM mitigation and ELM suppression closely depends on the LBO laser spot diameter.

  4. Simplified models for radiational losses calculating a tokamak plasma

    International Nuclear Information System (INIS)

    Arutiunov, A.B.; Krasheninnikov, S.I.; Prokhorov, D.Yu.

    1990-01-01

    To determine the magnitudes and profiles of radiational losses in a Tokamak plasma, particularly for high plasma densities, when formation of MARFE or detached-plasma takes place, it is necessary to know impurity distribution over the ionization states. Equations describing time evolution of this distribution are rather cumbersome, besides that, transport coefficients as well as rate constants of the processes involving complex ions are known nowadays with high degree of uncertainty, thus it is believed necessary to develop simplified, half-analytical models describing time evolution of the impurities analysis of physical processes taking place in a Tokamak plasma on the base of the experimental data. (author) 6 refs., 2 figs

  5. Near-wall effects in improved plasma confinement regimes in tokamak FT-2

    International Nuclear Information System (INIS)

    Budnikov, V.N.; D'yachenko, V.V.; Esipov, L.A.

    1997-01-01

    Transition to the regime of improved plasma confinement (H-mode) revealed in experiments on low hybrid heating in tokamak ft-2 is analyzed. Main attention is paid to processes, taking place in near-wall region. The data are correlated with results of experiments in large tokamaks

  6. Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions

    International Nuclear Information System (INIS)

    Sengoku, Seio

    1985-08-01

    Relation between plasma-wall interactions and confinement characteristics of a tokamak plasma with respect to both impurity and fuel particle controls is discussed. Following results are obtained from impurity control studies: (1) Ion sputtering is the dominant mechanism of impurity release in a steady state tokamak discharge. (2) By applying carbon coating on entire first wall of DIVA tokamak, dominant radiative region is concentrated more in boundary plasma resulting a hot peripheral plasma with cold boundary plasma. (3) A physical model of divertor functions about impurity control is empilically obtained. By a computer simulation based on above model with respect to divertor functions for JT-60 tokamak, it is found that the allowable electron temperature of the divertor plasma is not restricted by a condition that the impurity release due to ion sputtering does not increase continuously. (4) Dense and cold divertor plasma accompanied with strong remote radiative cooling was diagnosed along the magnetic field line in the simple poloidal divertor of DOUBLET III tokamak. Strong particle recycling region is found to be localized near the divertor plate. by and from particle control studies: (1) The INTOR scaling on energy confinement time is applicable to high density region when a core plasma is fueled directly by solid deuterium pellet injection in DOUBLET III tokamak. (2) As remarkably demonstrated by direct fueling with pellet injection, energy confinement characteristics can be improved at high density range by decreasing particle deposition at peripheral plasma in order to reduce plasma-wall interaction. (3) If the particle deposition at boundary layer is necessarily reduced, the electron temperature at the boundary or divertor region increases due to decrease of the particle recycling and the electron density there. (J.P.N.)

  7. Bayesian soft x-ray tomography and MHD mode analysis on HL-2A

    Science.gov (United States)

    Li, Dong; Liu, Yi; Svensson, J.; Liu, Y. Q.; Song, X. M.; Yu, L. M.; Mao, Rui; Fu, B. Z.; Deng, Wei; Yuan, B. S.; Ji, X. Q.; Xu, Yuan; Chen, Wei; Zhou, Yan; Yang, Q. W.; Duan, X. R.; Liu, Yong; HL-2A Team

    2016-03-01

    A Bayesian based tomography method using so-called Gaussian processes (GPs) for the emission model has been applied to the soft x-ray (SXR) diagnostics on HL-2A tokamak. To improve the accuracy of reconstructions, the standard GP is extended to a non-stationary version so that different smoothness between the plasma center and the edge can be taken into account in the algorithm. The uncertainty in the reconstruction arising from measurement errors and incapability can be fully analyzed by the usage of Bayesian probability theory. In this work, the SXR reconstructions by this non-stationary Gaussian processes tomography (NSGPT) method have been compared with the equilibrium magnetic flux surfaces, generally achieving a satisfactory agreement in terms of both shape and position. In addition, singular-value-decomposition (SVD) and Fast Fourier Transform (FFT) techniques have been applied for the analysis of SXR and magnetic diagnostics, in order to explore the spatial and temporal features of the saturated long-lived magnetohydrodynamics (MHD) instability induced by energetic particles during neutral beam injection (NBI) on HL-2A. The result shows that this ideal internal kink instability has a dominant m/n  =  1/1 mode structure along with a harmonics m/n  =  2/2, which are coupled near the q  =  1 surface with a rotation frequency of 12 kHz.

  8. Effects of Collisionality on the Nonlinear Characteristics of Boundary Turbulence and Blob/hole Transport in Tokamak Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.; Yasuaki, K., E-mail: lijq@energy.kyoto-u.ac.jp [Kyoto University, Kyoto (Japan); Cheng, J.; Longwen, Y.; Jiaqi, D. [Southwestern Institute of Physics, Chengdu (China)

    2012-09-15

    Full text: Blob/hole dynamics near tokamak separatrix is of striking importance in determining the boundary transport. Based on simulations using an extended 2-region (edge/SOL) fluid model, we found that blob/hole dynamics are sensitively influenced by the plasma collisionality, i.e., ion-electron and ion-neutral collisions. Namely, the holes are enhanced in highly collisional edge whereas the blobs are weakened at the SOL, causing larger particle convection. These blob/hole dynamics are closely correlated with potential dipoles. The trends are experimentally evidenced on the HL-2A tokamak. Moreover, as the neutral-ion collision increases, the blobs at the SOL tend to develop into streamers propagating outwards with reduced amplitude while the holes inwards are suppressed, showing a key role in nonlinear structure regulation and resultant transport suppression. Results suggest that adjusting the plasma collisionality by fueling, e.g., gas puffing, could serve as a method to nonlinearly select turbulent structures, i.e., blobs, holes or streamers, to access the control of boundary transport. (author)

  9. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1986-10-01

    It is shown that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. Gun injection raises the line-averaged density and peaks the density profile. Trapping of the gun-injected plasma is explainable in terms of a depolarization current mechanism. A model is developed which describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/ ∞ n -1 /sub b/T 3 /sub e/(α 0 /L) 2 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions

  10. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

    LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

    1988-04-01

    The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

  11. Numerical study for determining PF coil system parameters in MHD equilibrium of KT-2 tokamak plasma

    International Nuclear Information System (INIS)

    Ryu, J.; Hong, S.H.; Lee, K.W.; Hong, B.G.; In, S.R.; Kim, S.K.

    1995-01-01

    The KT-2 is a large-aspect-ratio medium-sized divertor tokamak in the conceptual design phase and planned to be operational in 1998 at the Korea Atomic Energy Research Institute (KAERI). Plasma equilibrium in tokamak can be acquired by controlling the current of poloidal field (PF) coils in appropriate geometries and positions. In this study, the authors have performed numerical calculations to achieve the various equilibrium conditions fitting given plasma shapes and satisfying PF current limitations. Usually an ideal magnetohydrodynamic (MHD) equation is used to obtain the equilibrium solution of tokamak plasma, and it is practical to take advantage of a numerical method in solving the MHD equation because it has nonlinear source terms. Two equilibrium codes have been applied to find a double-null configuration of free-boundary tokamak plasma in KT-2: one is of the authors' own developing and the other is a free-boundary tokamak equilibrium code (FBT) that has been used mainly for the verification of developed code's results. PF coil system parameters including their positions and currents are determined for the optimization of input power required when the specifications of KT-2 tokamak are met. Then, several sets of equilibrium conditions during the tokamak operation are found to observe the changes of poloidal field currents with the passing of operation time step, and the basic stability problems related with the magnetic field structure is also considered

  12. Poloidal rotation driven by electron cyclotron resonance wave in tokamak plasmas

    Directory of Open Access Journals (Sweden)

    Qing Zhou

    2017-10-01

    Full Text Available The poloidal electric filed, which is the drive field of poloidal rotation, has been observed and increases obviously after the injection of electron cyclotron resonance wave in HL-2A experiment, and the amplitude of the poloidal electric field is in the order of 103 V/m. Through theoretical analysis using Stringer rotation model, the observed poloidal electric field is of the same order as the theoretical calculation value. In addition, the magnetic pump damping which would damp the poloidal rotation is calculated numerically and the calculation results show that the closer to the core plasmas, the stronger the magnetic pump damping will be. Meanwhile, according to the value of the calculated magnetic pump damping, the threshold of the poloidal electric field which could overcome magnetic pump damping and drive poloidal rotation in tokamak plasmas is given out. Finally, the poloidal rotation velocity over time at different minor radius is studied theoretically.

  13. Plasma facing components design of KT-2 tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

  14. Numerical simulation of edge plasma in tokamak

    International Nuclear Information System (INIS)

    Chen Yiping; Qiu Lijian

    1996-02-01

    The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)

  15. Steady state operation of tokamaks. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2000-10-01

    The first IAEA Technical Committee Meeting (TCM) on Steady State Operation of Tokamaks was organized to discuss the operations of present long-pulse tokamaks (TRIAM-1M, TORE SUPRA, MT-7, HT-7M, HL-1M) and the plans for future steady-state tokamaks such as SST-1, CIEL, and HT-7U. This meeting, held from 13-15 October 1998, was hosted by the Academia Sinica Institute of Plasma Physics (ASIPP), Hefei, China. Participants from China, France, India, Japan, the Russian Federation, and the IAEA participated in the meeting. There were 18 individual presentations plus general discussions on many topics, including superconducting magnet systems, cryogenics, plasma position control, non-inductive current drive, auxiliary heating, plasma-wall interactions, high heat flux components, particle control, and data acquisition

  16. Asymmetry of edge plasma turbulence in biasing experiments on tokamak TF-2

    International Nuclear Information System (INIS)

    Budaev, V.P.

    1994-01-01

    It was observed in tokamaks the suppression of edge turbulence causes by setting a radial electric field at the plasma edge. The poloidal plasma rotation governed by this electric field is likely to result in changes in edge convention and poloidal asymmetry, however there is no experimental evidence about that of the experimental database concerning the biasing and conditions of edge plasma electrostatic turbulence excitation is not still complete. Also a relation between macroscopic convection and small-scale electrostatic turbulence have not yet revealed both in biasing and non biasing plasmas. In this paper results from biasing experiments carried on on ohmically heated tokamak TF-2 are presented. Changes in both equilibrium and fluctuated edge plasma parameters also convection and turbulence driven particle flux were demonstrated in probe measurements with biasing of electrode immersed within Last Closed Flux Surface (LCFS). Poloidal edge plasma structure and charge in asymmetry have demonstrated in the biasing experiments. (author). 6 refs, 4 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    Sarazin, Y

    2004-03-01

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

  18. 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].

  19. Studies on fundamental technologies for producing tokamak-plasma

    International Nuclear Information System (INIS)

    Matsuzaki, Yoshimi

    1987-10-01

    The report describes studies on fundamental technologies to produce tokamak-plasma of the JFT-2 and JFT-2M tokamaks. (1) In order to measure the particle number of residual gases, calibration methods of vacuum gauges have been developed. (2) Devices for a Taylor-type discharge cleaning (TDC), a glow discharge cleaning (GDC) and ECR discharge cleaning (ECR-DC) have been made and the cleaning effects have been investigated. In TDC the most effective plasma for cleaning is obtained in the plasma with 5 eV of electron temperature. GDC is effective in removing carbon impurities, but is less effective for removing oxygen impurities. ECR-DC has nearly the similar effect as TDC. The cleaning effect of these three types were studied by comparing the properties of resulting tokamak plasmas in the JFT-2M tokamak. (3) Experimental studies of pre-ionization showed as following results; A simple pre-ionization equipment as a hot-electron-gun and a J x B gun was effective in reducing breakdown voltage. An ordinary mode wave of the electron cyclotron frequency was very effective for pre-ionization. The RF power whose density is 3.6 x 10 -2 W/cm 3 produced plasma of an electron density of 5 x 10 11 cm -3 . In this case, it is possible to start up with negligible consumption of the magnetic flux caused by the plasma resistance. (4) Concerning to studies on plasma control, the following results were obtained; In order to obtain constant plasma current, a pulse forming network was constructed and sufficient constant plasma current was achieved. In applying an iso-flux method for measuring the plasma position, it is no problem practically to use only one loop-coil and one magnetic probe. (author)

  20. Dynamics and feedback control of plasma equilibrium position in a tokamak

    International Nuclear Information System (INIS)

    Burenko, O.

    1983-01-01

    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

  1. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

    1981-02-01

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  2. Pseudo-MHD ballooning modes in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1996-08-01

    The MHD description of a plasma is extended to allow electrons to have both fluid-like and adiabatic-regime responses within an instability eigenmode. In the resultant open-quotes pseudo-MHDclose quotes model, magnetic field line bending is reduced in the adiabatic electron regime. This makes possible a new class of ballooning-type, long parallel extent, MHD-like instabilities in tokamak plasmas for α > s 2 (2 7/3 /9) (r p /R 0 ) or-d√Β/dr > (2 1/6 /3)(s/ R 0q ), which is well below the ideal-MHD stability boundary. The marginally stable pressure profile is similar in both magnitude and shape to that observed in ohmically heated tokamak plasmas

  3. Plasma position control in TCABR Tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

    1998-01-01

    The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

  4. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1985-01-01

    The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

  5. Edge plasma physical investigations of tokamak plasmas in CRIP

    International Nuclear Information System (INIS)

    Bakos, J.; Ignacz, P.; Koltai, L.; Paszti, F.; Petravich, G.; Szigeti, J.; Zoletnik, S.

    1988-01-01

    The results of the measurements performed in the field of thermonuclear high temperature plasma physics in CRIP (Hungary) are summarized. In the field of the edge plasma physics solid probes were used to test the external zone of plasma edges, and atom beams and balls were used to investigate both the external and internal zones. The plasma density distribution was measured by laser blow-off technics, using Na atoms, which are evaporated by laser pulses. The excitation of Na atom ball by tokamak plasma gives information on the status of the plasma edge. The toroidal asymmetry of particle transport in tokamak plasma was measured by erosion probes. The evaporated and transported impurities were collected on an other part of the plasma edge and were analyzed by SIMS and Rutherford backscattering. The interactions in plasma near the limiter were investigated by a special limiter with implemented probes. Recycling and charge exchange processes were measured. Disruption phenomena of tokamak plasma were analyzed and a special kind of disruptions, 'soft disruptions' and the related preliminary perturbations were discovered. (D.Gy.) 10 figs

  6. Orbit effects on impurity transport in a rotating tokamak plasma

    International Nuclear Information System (INIS)

    Wong, K.L.; Cheng, C.Z.

    1988-05-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster with a higher bounce frequency, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle orbits near the surface of a rotating tokamak are also analyzed. Orbit effects indicate that more impurities can penetrate into a plasma rotating with counter-beam injection. Particle simulation is carried out with realistic experimental parameters and the results are in qualitative agreement with some experimental observations in the Tokamak Fusion Test Reactor (TFTR). 19 refs., 15 figs

  7. Intelligent plasma operation and control system for HL-2M

    International Nuclear Information System (INIS)

    Xia, F.; Chen, L.Y.; Wang, C.; Zhang, G.; Song, X.M.; Song, X.; Pan, L.; Zhao, L.; Pan, W.; Lan, J.T.

    2015-01-01

    Full text of publication follows. The Intelligent Plasma Operation and Control System (IPOCS) for HL-2M is under construction based on the current actual situation of HL-2A control system in SWIP. The purpose of IPOCS is to replace the routine processes before, during and post discharge with the automatic process algorithms and give various information, for example, status, warning, reason, instruction, suggestion, physics phenomenon etc. to staff who can access the intranet. In this case, the core research objectives can be focused on and the related operations can be carried out referring to the information. The ultimate goal of IPOCS is to improve the efficiency of plant operation and control and physics research for HL-2M and for the fusion reactor in the future. There are three layers in IPOCS, which are named the information collection layer, the data processing layer and the presentation layer respectively. Information collection layer consists of data acquisition system, EPICS system, Audio and Video system, discharge scheduling system, data storage system etc. All raw data are collected and stored in this layer. The data processing layer is the core of IPOCS. Most algorithms are executed here. The basic computation platform is based on cluster (56 cores at present) and Parallel Computing Toolbox provided by Matlab. The physics database for SWIP based on HDF5 and offline EFIT also locate at this layer. All the operations can be finished during the time interval between two shots. The last layer is to present the information from the former two layers. The typical hardware including the large display screen system in the control room, the voice broadcasting system, personal monitors and smart phones, etc.. Several applications has being developed such as Control System Studio (CSS) in SWIP version, WebOPI, Automatic Alarm Display system and so on. IPOCS for HL-2M is at the very beginning phase at present. With more and more systems and algorithms got

  8. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  9. Plasma transport in a compact ignition tokamak

    International Nuclear Information System (INIS)

    Singer, C.E.; Ku, L.P; Bateman, G.

    1987-02-01

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

  10. Scaling for scrape-off layer plasma in tokamak

    International Nuclear Information System (INIS)

    Shimomura, Yasuo; Maeda, Hikosuke; Kimura, Haruyuki; Azumi, Masashi; Odajima, Kazuo

    1977-12-01

    Scaling for a scrape-off layer plasma in a tokamak is obtained by using DIVA (JFT-2a). The scaling gives the average electron temperature, the width and the mean electron density of the scrape-off layer. The temperature at the edge will be high in a future large tokamak with a small energy-loss by charge-exchange and radiation. The scrape-off layer plasma can easily shield the impurity influx from the wall. The fuel, however, can easily penetrate into the main plasma. (auth.)

  11. Study on the bywalled plasma near the FT-2 tokamak limiter

    International Nuclear Information System (INIS)

    Budnikov, V.N.; Gerasimenko, G.V.; Esipov, L.A.; Levitskij, A.N.; Sakharov, I.E.; Shatalin, S.V.

    1987-01-01

    Results od investigation into the bywalled plasma near the FT-2 tokamak limiter, conducted using the ternary probe method are presented. Plasma charged particle concentration radial profiles, results of measuring the electron temperature and azimuthal electric field are presented. It is shown, that at l > 2mm, where l is the distance up to the limiter, concentration practically does not depend on the longitudinal coordinate and near the limiter (l < 20 mm) a sharp drop of plasma charged particle concentration measured values (∼ 2-fold) is observed. Neoclassical mechanism of filling the depletion area (the probe shade) is confirmed

  12. Plasma Equilibrium Control in Nuclear Fusion Devices 2. Plasma Control in Magnetic Confinement Devices 2.1 Plasma Control in Tokamaks

    Science.gov (United States)

    Fukuda, Takeshi

    The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.

  13. Plasma-neutral gas interaction in a tokamak divertor: effects of hydrogen molecules and plasma recombination

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Pigarov, A.Yu.; Soboleva, T.K.; Sigmar, D.J.

    1997-01-01

    We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 x 10 -10 cm 3 /s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a 'gas box' divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (orig.)

  14. 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.

  15. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-01-01

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

  16. Full Tokamak discharge simulation and kinetic plasma profile control for ITER

    International Nuclear Information System (INIS)

    Hee Kim, S.

    2009-10-01

    Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

  17. 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).

  18. Preliminary design of HL-2A discharge control system

    International Nuclear Information System (INIS)

    Jiang Chao; Song Xianming; Li Qiang

    2001-01-01

    HL-2A Discharge Control System consists of one or more VXI work stations so as to compose an all digital control system. The DCS are used to measure and control the poloidal coils, the main tasks of the poloidal coils are exploding, keeping and controlling the current of plasma. These coils explode plasma and keep it in the determined position

  19. Plasma internal inductance dynamics in a tokamak

    International Nuclear Information System (INIS)

    Romero, J.A.

    2010-01-01

    A lumped parameter model for tokamak plasma current and inductance time evolution as a function of plasma resistance, non-inductive current drive sources and boundary voltage or poloidal field coil current drive is presented. The model includes a novel formulation leading to exact equations for internal inductance and plasma current dynamics. Having in mind its application in a tokamak inductive control system, the model is expressed in state space form, the preferred choice for the design of control systems using modern control systems theory. The choice of system states allows many interesting physical quantities such as plasma current, inductance, magnetic energy, and resistive and inductive fluxes be made available as output equations. The model is derived from energy conservation theorem, and flux balance theorems, together with a first order approximation for flux diffusion dynamics. The validity of this approximation has been checked using experimental data from JET showing an excellent agreement.

  20. Plasma edge physics in an actively cooled tokamak

    International Nuclear Information System (INIS)

    Gunn, J.P.; Adamek, A.; Boucher, C.

    2005-01-01

    Tore Supra is a large tokamak with a plasma of circular cross section (major radius 2.4 m and minor radius 0.72 m) lying on a toroidal limiter. Tore Supra's main mission is the development of technology to inject up to 25 MW of microwave heating power and extract it continuously for up to 1000 s in steady state without uncontrolled overheating of, or outgassing from, plasma-facing components. The entire first wall of the tokamak is actively cooled by a high pressure water loop and special carbon fiber composite materials have been designed to handle power fluxes up to 10 MW/m 2 . The edge plasma on open magnetic flux surfaces that intersect solid objects plays an important role in the overall behaviour of the plasma. The transport of sputtered impurity ions and the fueling of the core plasma are largely governed by edge plasma density, temperature, and flow profiles. Measurements of these quantities are becoming more reliable and frequent in many tokamaks, and it has become clear that we do not understand them very well. Classical two-dimensional fluid modelling fails to reproduce many aspects of the experimental observations such as the significant thickness of the edge plasma, and the near-sonic flows that occur where none should be expected. It is suspected that plasma turbulence is responsible for these anomalies. In the Tore Supra tokamak, various kinds of Langmuir probes are used to characterize the edge plasma. We will present original measurements that demonstrate the universality of many phenomena that have been observed in X-point divertor tokamaks, especially concerning the ion flows. As in the JET tokamak, surprisingly large values of parallel Mach number are measured midway between the two strike zones, where one would expect to find nearly stagnant plasma if the particle source were poloidally uniform. We will present results of a novel experiment that provides evidence for a poloidally localized particle and energy source on the outboard midplane of

  1. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-08-01

    Dynamic control of the plasma position within the torus of a TOKAMAK fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. (auth)

  2. Beat-wave excitation and current driven in tokamak plasma. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, B F [Plasma physics Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    Wave heating current drive in tokamaks is a growing subject in the plasma physics literature. For current drive in tokamaks by electromagnetic waves, different methods have been proposed recently. One of the promising schemes for current drive remains the beat wave scheme. This technique employs two CO- or counterpropagating monochromatic laser beams (or microwaves) whose frequency difference matches the plasma frequency, while the wave number difference (or sum, in the case of counterpropagating) determine the wave number of the resulting plasma beat wave. In this work, the basic analysis of a beat wave current drive scheme in which collinear waves are used is discussed. by assuming a Gaussian profile for the amplitude of these pump waves, the amplitudes of the longitudinal and radial fields of the beat wave due to the nonlinear wave interactions have been calculated. Besides, the transfer of momentum flux that accompanies the transfer of wave action in beat-wave scattering will be used to drive the toroidal radial currents in tokamaks. self-generated magnetic fields due to those currents were also calculated. 1 fig.

  3. Edge plasma diagnostics on Tore Supra tokamak

    International Nuclear Information System (INIS)

    Fujita, Junji

    1991-01-01

    From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)

  4. A Midsize Tokamak As Fast Track To Burning Plasmas

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2010-01-01

    This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain ((ge) 10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.

  5. Snake perturbations during pellet injection and LHCD in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Liu Yi; Qiu Xiaoming; Dong Yunbo; Zhong Yunzhe; Fu Bingzhong; Jiafu Dong Yong Liu

    2005-01-01

    Excitation of snake perturbations has been observed in the core region of pellet-fuelled HL-1M plasmas when the pellets cross surface with q value 1. Through measurements of plasma q profile by means of multi-exposures with CCD camera during pellet ablation, and investigation on pellet ablation process, possible mechanisms for the formation of snake oscillation are discussed. In addition, a large, long-lived snake-like oscillation is frequently observed in lower hybrid current driven discharge in which the sawtooth has been stabilized at early times. There is evidence that such a perturbation is due to impurity accumulation during sawtooth-stabilization, and the good performance with peaking profiles after LHCD is limited by magnetohydrodynamic (MHD) instabilities including sawtooth and snake activities in HL-1M plasma. (author)

  6. An overview on plasma disruption mitigation and avoidance in tokamak

    International Nuclear Information System (INIS)

    He Kaihui; Pan Chuanhong; Feng Kaiming

    2002-01-01

    Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

  7. A complex probe for tokamak plasma edge conditions

    International Nuclear Information System (INIS)

    Castro, R.M. de; Silva, R.P. da; Heller, M.V.A.P.; Caldas, I.L.; Nascimento, I.C.; Degasperi, F.T.

    1995-01-01

    The study of the physical processes that occur in the plasma edge of tokamak machines has recently grown due to the evidence that these processes influence those that occur in the center of the plasma column. Experimental studies show the existence of a strong level of fluctuations in the plasma edge. The results of these studies indicate that these fluctuations enhance particle and energy transport and degrade the confinement. In order to investigate these processes in the plasma edge of the TBR-1 Tokamak, a Langmuir probe array, a triple and a set of magnetic probes have been designed and constructed. With this set probes the mean and fluctuation values of the magnetic field were detected and correlated with the fluctuating parameters obtained with the electrostatic probes. (author). 7 refs., 5 figs

  8. Impedance of an intense plasma-cathode electron source for tokamak startup

    Science.gov (United States)

    Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.

    2016-05-01

    An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

  9. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    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.

    2004-03-01

    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

  10. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1986-01-01

    It is shown that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. Gun injection raises the line-averaged density and peaks the density profile. Trapping of the gun-injected plasma is explainable in terms of a depolarization current mechanism

  11. Development of a visualized software for tokamak experiment data processing

    International Nuclear Information System (INIS)

    Cao Jianyong; Ding Xuantong; Luo Cuiwen

    2004-01-01

    With the VBA programming in Microsoft Excel, the authors have developed a post-processing software of experimental data in tokamak. The standard formal data in the HL-1M and HL-2A tokamaks can be read, displayed in Excel, and transmitted directly into the MATLAB workspace, for displaying pictures in MATLAB with the software. The authors have also developed data post-processing software in MATLAB environment, which can read standard format data, display picture, supply visual graphical user interface and provide part of advanced signal processing ability

  12. Role of advanced refuelling and heating on edge Reynolds stress-induced poloidal flow in HL-1M

    International Nuclear Information System (INIS)

    Hong Wenyu; Wang Enyao; Li Qiang; Cao Jianyong; Yan Longwen

    2002-01-01

    The radial profile of electrostatic Reynolds stress, plasma poloidal rotations, radial and poloidal electric fields have been measured in the plasma boundary region of the HL-1M tokamak using a multi-array of Mach/Langmuir probes. In the experiments of ohmic discharge, lower hybrid current drive, supersonic molecular beam injection (SMBI) and multi-shot pellet injection, the correlation between the Reynolds stress and poloidal flow in the edge plasma is presented. The radial profile changes of the Reynolds stress and poloidal flow velocity V pol with lower hybrid wave injection power and SMBI injection are obtained. The results indicate that the sheared poloidal flow can be generated in tokamak plasma due to the radially varying Reynolds stress

  13. Boundary Plasma Turbulence Simulations for Tokamaks

    International Nuclear Information System (INIS)

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

    2008-05-01

    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 e ; T 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

  14. Development of the 'JFT-2' tokamak plasma position control system

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.

    1980-01-01

    Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)

  15. Impurity screening in high density plasmas in tokamaks with a limiter configuration

    International Nuclear Information System (INIS)

    Ferro, C.; Zanino, R.

    1992-01-01

    Impurity screening in high density plasmas in tokamaks with a limiter configuration is investigated by means of a simple semi-analytical model. An iterative scheme is devised, in order to determine self-consistently the values of scrape-off layer thickness, edge electron density and temperature, and main plasma contamination parameter Z eff , as a function of given average electron density and temperature in the main plasma and given input power. The model is applied to the poloidal limiter case of the Frascati Tokamak Upgrade, and results are compared with experimental data. A reasonable agreement between the trends is found, emphasizing the importance of a high edge plasma density for obtaining a clean main plasma in limiter tokamaks. (orig.)

  16. Power supply for plasma generator of HL-1M neutral beam injector

    International Nuclear Information System (INIS)

    Wang Detai; Qian Jiamei; Lei Guangjiu; Shun Mengda; Jiang Shaofeng; Wang Enyao; Lu Xuejun; Yang Tiehai; Wang Xuehua; Zhao Zhimin; Hao Ming; Huang Jianrong; Yu Yanqiu; Cheng Baoqiang; Wu Zhige; Sheng Ning; Hu Qingtao

    1999-01-01

    The diagram of the HL-1M Neutral Beam Injector (NBI) and the power supply (PS) system is shown. The NBI consists of ion source, beam line and power supply system etc. The ion source includes plasma generator and three-electrode extraction system. The power supply for plasma generator consists of a filament PS, an arc PS and gas valve PS. Testing has shown that the PS for plasma generator of the HL-1M NBI has excellent stability and obtain good plasma heating effect

  17. Technology and plasma-materials interaction processes of tokamak disruptions

    International Nuclear Information System (INIS)

    McGrath, R.T.; Kellman, A.G.

    1992-01-01

    A workshop on the technology and plasma-materials interaction processes of tokamak disruptions was held April 3, 1992 in Monterey, California, as a satellite meeting of the 10th International Conference on Plasma-Surface Interactions. The objective was to bring together researchers working on disruption measurements in operating tokamaks, those performing disruption simulation experiments using pulsed plasma gun, electron beam and laser systems, and computational physicists attempting to model the evolution and plasma-materials interaction processes of tokamak disruptions. This is a brief report on the workshop. 4 refs

  18. Neutron measurement techniques for tokamak plasmas

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1994-01-01

    The present article reviews the neutron measurement techniques that are currently being applied to the study of tokamak plasmas. The range of neutron energies of primary interest is limited to narrow bands around 2.5 and 14 MeV, and the variety of measurements that can be made for plasma diagnostic purposes is also restricted. To characterize the plasma as a neutron source, it is necessary only to measure the total neutron emission, the relative neutron emissivity as a function of position throughout the plasma, and the energy spectra of the emitted neutrons. In principle, such measurements might be expected to be relatively easy. That this is not the case is, in part, attributable to practical problems of accessibility to a harsh environment but is mostly a consequence of the time-scale on which the measurements have to be made and of the wide range of neutron emission intensities that have to be covered: for tokamak studies, the time-scale is of the order of 1 to 100 ms and the neutron intensity ranges from 10 12 to 10 19 s -1 . (author)

  19. Digital control of plasma position in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

    2002-03-01

    Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

  20. Magnetohydrodynamic stability of tokamak edge plasmas

    International Nuclear Information System (INIS)

    Connor, J.W.; Hastie, R.J.; Wilson, H.R.; Miller, R.L.

    1998-01-01

    A new formalism for analyzing the magnetohydrodynamic stability of a limiter tokamak edge plasma is developed. Two radially localized, high toroidal mode number n instabilities are studied in detail: a peeling mode and an edge ballooning mode. The peeling mode, driven by edge current density and stabilized by edge pressure gradient, has features which are consistent with several properties of tokamak behavior in the high confinement open-quotes Hclose quotes-mode of operation, and edge localized modes (or ELMs) in particular. The edge ballooning mode, driven by the pressure gradient, is identified; this penetrates ∼n 1/3 rational surfaces into the plasma (rather than ∼n 1/2 , expected from conventional ballooning mode theory). Furthermore, there exists a coupling between these two modes and this coupling provides a picture of the ELM cycle

  1. 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

  2. Analysis of tokamak plasma confinement modes using the fast

    Indian Academy of Sciences (India)

    The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...

  3. Interaction of a spheromak-like compact toroid with a high beta spherical tokamak plasma

    International Nuclear Information System (INIS)

    Hwang, D.Q.; McLean, H.S.; Baker, K.L.; Evans, R.W.; Horton, R.D.; Terry, S.D.; Howard, S.; Schmidt, G.L.

    2000-01-01

    Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak. (author)

  4. Self-organized ignition of a tokamak plasma

    International Nuclear Information System (INIS)

    Schoepf, K.

    2007-01-01

    The continuous progress in the attainment of plasma parameters required for establishing nuclear fusion in magnetically confined plasmas as well as the prospect of feasible steady-state operation has instigated the interest in the physics of burning plasmas [1]. Aside from the required plasma current drive, fusion energy production with tokamaks demands particular attention to confinement and fuelling regimes in order to maintain the plasma density n and temperature T at favourable values matching with specific requirements such as the triple product nτ E T, where τ E represents the plasma energy confinement time. The identification of state and parameter space regions capable of ignited fusion plasma operation is evidently crucial if significant energy gains are to be realized over longer periods. Examining the time-evolving state of tokamak fusion plasma in a parameter space spanned by the densities of plasma constituents and their temperatures has led to the formation of an ignition criterion [2] fundamentally different from the commonly used static patterns. The incorporation of non-stationary particle and energy balances into the analysis here, the application of a 'soft' Troyon beta limit [3], the consideration of actual fusion power deposition [4,5] and its effect of reducing τ E are seen to significantly influence the fusion burn dynamics and to shape the ignition conditions. The presented investigation refers to a somewhat upgraded (to achieve ignition) ITER-like tokamak plasma and uses volume averages of locally varying quantities and processes. The resulting ignition criterion accounts for the dynamic evolution of a reacting plasma controlled by heating and fuel feeding. Interestingly, also self-organized ignition can be observed: a fusion plasma possessing a density and temperature above a distinct separatrix in the considered parameter phase space is seen to evolve - without external heating and hence practically by itself - towards an ignited

  5. Experimental investigation of molecular beam injection in HL-1 tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Diao Guangyao; Wang Zhanhe; Deng Huichen; Luo Junlin; Duan Xuru; Cui Zhengying

    1993-07-01

    A new method of gas puffing is presented. The molecular beam, formed by high pressure deuterium gas through Larval nozzle and skimmer slit, is injected into the HL-1 vacuum vessel. The deuterium molecular current from the nozzle passing through the skimmer is about 3 x 10 20 /s. At the line average electron density of 5.2 x 10 19 m -3 , the beam velocity is about 100 m/s. As the plasma density and temperature increasing, the influxes of deuterium particles attenuate quickly. When the molecular beam injection (MBI) just returned to normal gas puffing, the D α emission rapidly decreases, meanwhile, the particles move toward plasma center, the electron density is continuously peaking. The line average electron density rising lasts 45 ms. The thermal energy of plasma and confinement time for particles and energy are also increasing. the MBI is a direct and efficient gas fuelling mode, and the injected particles can reach to inside about 8 cm of plasma and q ≅ 2 confinement region. Its efficiency of injection is about 50%. After the MBI, the particle recycling coefficient R on the wall is 0.6 which is 10% lower than that of normal gas puffing

  6. Characteristics of disruptive plasma current decay in the HT-2 tokamak

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Takeuchi, Kazuhiro; Otsuka, Michio

    1993-01-01

    Motions of plasma current channel and time evolutions of eddy current distribution on the vacuum vessel during disruptive plasma current decay were studied experimentally in the Hitachi tokamak HT-2. The plasmas are vertically elongated and circularly shaped plasmas. A disruptive plasma current decay has three phases. During the first phase, a large displacement of the plasma position without plasma current decay is observed. Rapid plasma current decay is observed during the second phase and the decay rate is roughly constant with time. The eddy current distribution is like that due to the shell effect which creates a poloidal field to reduce the plasma displacement. During the third phase, the plasma current decays exponentially. The second phase is observed in slightly elongated and high plasma current (> 20 kA) circularly shaped plasmas. The plasma current decay rates in the second phase depend on the plasma cross sectional shape, but they do not in the third phase. The magnetic axis moves from the plasma area to the vacuum vessel wall between the second and third phases. (author)

  7. MHD Effects of a Ferritic Wall on Tokamak Plasmas

    Science.gov (United States)

    Hughes, Paul E.

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

  8. 2D full-wave simulation of waves in space and tokamak plasmas

    Directory of Open Access Journals (Sweden)

    Kim Eun-Hwa

    2017-01-01

    Full Text Available Simulation results using a 2D full-wave code (FW2D for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF waves in the scape-off layer (SOL of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  9. 2D full-wave simulation of waves in space and tokamak plasmas

    Science.gov (United States)

    Kim, Eun-Hwa; Bertelli, Nicola; Johnson, Jay; Valeo, Ernest; Hosea, Joel

    2017-10-01

    Simulation results using a 2D full-wave code (FW2D) for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry) and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC) waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF) waves in the scape-off layer (SOL) of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  10. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Energy Technology Data Exchange (ETDEWEB)

    Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Kocman, J.; Grover, O. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Krbec, J.; Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, CZ-182 21 Prague (Czech Republic)

    2015-10-15

    Graphical abstract: * Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes.* Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform.* More than 20% plasma life prolongation with plasma position control in feedback mode. - Highlights: • Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes. • Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform. • More than 20% plasma life prolongation with plasma position control in feedback mode. - Abstract: The GOLEM tokamak at the Czech Technical University has been established as an educational tokamak device for domestic and foreign students. Remote participation in the scope of several laboratory practices, plasma physics schools and workshops has been successfully performed from abroad. A new enhancement allowing understandable remote control of vertical plasma position in two modes (i) predefined and (ii) feedback control is presented. It allows to drive the current in the stabilization coils in any time-dependent scenario, which can include as a parameter the actual plasma position measured by magnetic diagnostics. Arbitrary movement of the plasma column in a vertical direction, stabilization of the plasma column in the center of the tokamak vessel as well as prolongation/shortening of plasma life according to the remotely defined request are demonstrated.

  11. Bifurcated states of a rotating tokamak plasma in the presence of a static error-field

    International Nuclear Information System (INIS)

    Fitzpatrick, R.

    1998-01-01

    The bifurcated states of a rotating tokamak plasma in the presence of a static, resonant, error-field are strongly analogous to the bifurcated states of a conventional induction motor. The two plasma states are the open-quotes unreconnectedclose quotes state, in which the plasma rotates and error-field-driven magnetic reconnection is suppressed, and the open-quotes fully reconnectedclose quotes state, in which the plasma rotation at the rational surface is arrested and driven magnetic reconnection proceeds without hindrance. The response regime of a rotating tokamak plasma in the vicinity of the rational surface to a static, resonant, error-field is determined by three parameters: the normalized plasma viscosity, P, the normalized plasma rotation, Q 0 , and the normalized plasma resistivity, R. There are 11 distinguishable response regimes. The extents of these regimes are calculated in P endash Q 0 endash R space. In addition, an expression for the critical error-field amplitude required to trigger a bifurcation from the open-quotes unreconnectedclose quotes to the open-quotes fully reconnectedclose quotes state is obtained in each regime. The appropriate response regime for low-density, ohmically heated, tokamak plasmas is found to be the nonlinear constant-ψ regime for small tokamaks, and the linear constant-ψ regime for large tokamaks. The critical error-field amplitude required to trigger error-field-driven magnetic reconnection in such plasmas is a rapidly decreasing function of machine size, indicating that particular care may be needed to be taken to reduce resonant error-fields in a reactor-sized tokamak. copyright 1998 American Institute of Physics

  12. Increase in beta limit in tokamak plasmas

    International Nuclear Information System (INIS)

    Kamada, Yutaka

    2003-01-01

    This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)

  13. Coherent structures in tokamak plasmas workshop: Proceedings

    International Nuclear Information System (INIS)

    Koniges, A.E.; Craddock, G.G.

    1992-08-01

    Coherent structures have the potential to impact a variety of theoretical and experimental aspects of tokamak plasma confinement. This includes the basic processes controlling plasma transport, propagation and efficiency of external mechanisms such as wave heating and the accuracy of plasma diagnostics. While the role of coherent structures in fluid dynamics is better understood, this is a new topic for consideration by plasma physicists. This informal workshop arose out of the need to identify the magnitude of structures in tokamaks and in doing so, to bring together for the first time the surprisingly large number of plasma researchers currently involved in work relating to coherent structures. The primary purpose of the workshop, in addition to the dissemination of information, was to develop formal and informal collaborations, set the stage for future formation of a coherent structures working group or focus area under the heading of the Tokamak Transport Task Force, and to evaluate the need for future workshops on coherent structures. The workshop was concentrated in four basic areas with a keynote talk in each area as well as 10 additional presentations. The issues of discussion in each of these areas was as follows: Theory - Develop a definition of structures and coherent as it applies to plasmas. Experiment - Review current experiments looking for structures in tokamaks, discuss experimental procedures for finding structures, discuss new experiments and techniques. Fluids - Determine how best to utilize the resource of information available from the fluids community both on the theoretical and experimental issues pertaining to coherent structures in plasmas. Computation - Discuss computational aspects of studying coherent structures in plasmas as they relate to both experimental detection and theoretical modeling

  14. Generation of plasma rotation by ICRH in tokamaks

    International Nuclear Information System (INIS)

    Chang, C.; Phillips, C.K.; White, R.B.; Zweben, S.; Bonoli, P.T.; Rice, J.; Greenwald, M.; Grassie, J.S. de

    2001-01-01

    A physical mechanism to generate plasma rotation by ICRH is presented in a tokamak geometry. By breaking the omnigenity of resonant ion orbits, ICRH can induce a non-ambipolar minor-radial flow of resonant ions. This induces a return current j p r in the plasma, which then drives plasma rotation through the j p r xB force. It is estimated that the fast-wave power in the present-day tokamak experiments can be strong enough to give a significant modification to plasma rotation. (author)

  15. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

  16. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  17. Low temperature plasma near a tokamak reactor limiter

    International Nuclear Information System (INIS)

    Braams, B.J.; Singer, C.E.

    1985-01-01

    Analytic and two-dimensional computational solutions for the plasma parameters near a toroidally symmetric limiter are illustrated for the projected parameters of a Tokamak Fusion Core Experiment (TFCX). The temperature near the limiter plate is below 20 eV, except when the density 10 cm inside the limiter contact is 8 x 10 13 cm -3 or less and the thermal diffusivity in the edge region is 2 x 10 4 cm 2 /s or less. Extrapolation of recent experimental data suggests that neither of these conditions is likely to be met near ignition in TFCX, so a low plasma temperature near the limiter should be considered a likely possibility

  18. Minimum dimension of an ITER like Tokamak with a given Q

    Energy Technology Data Exchange (ETDEWEB)

    Johner, J

    2004-07-01

    The minimum dimension of an ITER like tokamak with a given amplification factor Q is calculated for two values of the maximum magnetic field in the superconducting toroidal field coils. For ITERH-98P(y,2) scaling of the energy confinement time, it is shown that for a sufficiently large tokamak, the maximum Q is obtained for the operating point situated both at the maximum density and at the minimum margin with respect to the H-L transition. We have shown that increasing the maximum magnetic field in the toroidal field coils from the present 11.8 T to 16 T would result in a strong reduction of the machine size but has practically no effect on the fusion power. Values obtained for {beta}{sub N} are found to be below 2. Peak fluxes on the divertor plates with an ITER like divertor and a multi-machine expression for the power radiated in the plasma mantle, are below 10 MW/m{sup 2}.

  19. Snake perturbations during pellet injection and LHCD in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Liu Yi; Qiu Xiaoming; Dong Yunbo; Guo Gangcheng; Xiao Zhengui; Zhong Yunzhe; Zheng Yinjia; Fu Bingzhong; Dong Jiafu; Liu Yong; Wang Enyao

    2004-01-01

    Excitation of snake perturbations has been observed in the core region of pellet-fuelled HL-1M plasmas when the pellets cross the surface with a q value of 1. It is observed that the snake oscillations have an m = 1, n = 1 helicity with quite a long lifetime. A detailed comparison has been made between the locations of the q = 1 surface and the snake oscillation. Through measurements of the plasma q-profile by means of multi-exposures with a CCD camera during pellet ablation, and investigation of the pellet ablation process, possible mechanisms for the formation of the snake oscillation are discussed. In addition, a large, long-lived snake-like oscillation is frequently observed in lower-hybrid current driven (LHCD) discharge in which the sawtooth has been stabilized early in the discharge. There is evidence that such a perturbation is due to impurity accumulation during sawtooth-stabilization, and good performance with peaking profiles after LHCD is limited by magnetohydrodynamic instabilities including sawtooth and snake activities in HL-1M plasmas

  20. Behaviour of metallic droplets in a tokamak plasma

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

    1989-01-01

    Micrometre sized tantalum droplets were injected into a tokamak plasma by a controllable arcing gun located behind the wall. The trajectories of the ablating particles were photographed by a high speed camera. Various possible mechanisms which may explain the observed curvature of the particle paths are discussed. The migration of the ablated material in the tokamak was studied by post-mortem analysis of collector probes and limiters. (author). Letter-to-the-editor. 12 refs, 9 figs

  1. Simulations of fast component and slow component of SMBI on HL-2A tokamak

    International Nuclear Information System (INIS)

    Shi Yong-Fu; Wang Zhan-Hui; Sun Ai-Ping; Yu De-Liang; Xu Min; Ren Qi-Long; Guo Wen-Feng

    2017-01-01

    It is very important to improve the penetration depth and fueling efficiency of supersonic molecular beam injection (SMBI) especially for the next generation fusion devices such as ITER. Two components, a fast component (FC) and a slow component (SC), have been observed in the HL-2A SMBI experiments for several years, and the FC can penetrate much more deeply than the common SMBIs which draws a great deal of attention for a better fueling method. It is the first time to the FC and SC of SMBI have been simulated and interpreted in theory and simulation in this paper with the trans-neut module of the BOUT++ code. The simulation results of the FC and SC are clear and distinguishable in the same way as the observation in experiment. For the major mechanism of the FC and SC, it is found that although the difference in the injection velocity has some effect on the penetration depth difference between the FC and SC, it is mainly caused by the self-blocking effect of the first ionized SMB. We also discuss the influence of the initial plasma density on the FC and SC, and the variation of the SC penetration depth with its injection velocity. (paper)

  2. Rippling modes in the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Carreras, B.A.; Callen, J.D.; Gaffney, P.W.; Hicks, H.R.

    1982-02-01

    A promising resistive magnetohydrodynamic candidate for the underlying cause of turbulence in the edge of a tokamak plasma is the rippling instability. In this paper we develop a computational model for these modes in the cylindrical tokamak approximation and explore the linear growth and single-helicity quasi-linear saturation phases of the rippling modes for parameters appropriate to the edge of a tokamak plasma. Large parallel heat conduction does not stabilize these modes; it only reduces their growth rate by a factor scaling as k/sub parallel//sup -4/3/. Nonlinearly, individual rippling modes are found to saturate by quasi-linear flattening of the resistivity profile. The saturated amplitude of the modes scales as m/sup -1/, and the radial extent of these modes grows linearly with time due to radial Vector E x Vector B 0 convection. This evolution is found to be terminated by parallel heat conduction

  3. Rippling modes in the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Carreras, B.A.; Gaffney, P.W.; Hicks, H.R.; Callan, J.D.

    1982-01-01

    A promising resistive magnetohydrodynamic candidate for the underlying cause of turbulence in the edge of a tokamak plasma is the rippling instability. In this paper a computational model for these modes in the cylindrical tokamak approximation was developed and the linear growth and single-helicity quasi-linear saturation phases of the rippling modes for parameters appropriate to the edge of a tokamak plasma were explored. Large parallel heat conduction does not stabilize these modes; it only reduces their growth rate by a factor sacling as K/sup -4/3//sub parallel/. Nonlinearly, individual rippling modes are found to saturate by quasi-linear flattening of the resistivity profile. The saturated amplitude of the modes scales as m -1 , and the radial extent of these modes grows linearly with time due to radial E x B 0 convection. This evolution is found to be terminated by parallel heat conduction

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

    International Nuclear Information System (INIS)

    Amrollahi, R.

    1997-01-01

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

  5. Cleaning of the first mirrors and diagnostic windows by YAG laser on HL-2A

    International Nuclear Information System (INIS)

    Zhou, Y; Zheng, L; Li, Y G; Li, L C; Jiao, Y M; Gao, H; Zhao, G

    2009-01-01

    A laser cleaning system for HL-2A tokamak first mirrors and diagnostic windows has been developed recently. A detailed description of the laser cleaning procedure is presented. The optical transmission performance measured before and after the laser cleaning of the impurity film deposited on the optical elements is investigated. HL-2A deposited layers on metal mirrors and glass windows with thicknesses of about 1 and 4 μm, respectively, are clearly removed by irradiation with a single pulse of a Q-switched Nd:YAG laser with energy density of 0.4 and 2.8 J cm -2 , respectively. The feasibility of cleaning ECE windows is demonstrated. A cleaning time of about 5 min is suitable for application in fusion devices. The comparison of results obtained at different laser wavelengths shows that there is a greater probability of damage to the metallic mirror surface with a short laser wavelength than with longer wavelength.

  6. Plasma diagnostics on large tokamaks

    International Nuclear Information System (INIS)

    Orlinskij, D.V.; Magyar, G.

    1988-01-01

    The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs

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

    International Nuclear Information System (INIS)

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

    2009-01-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  8. The infrared camera system on the HL-2A tokamak device

    International Nuclear Information System (INIS)

    Li Wei; Lu Jie; Yi Ping

    2009-04-01

    In order to measure and analyze the heat flux on the divertor plate under different discharge conditions, an infrared camera diagnostic system for HL-2A Device has been developed. The infrared camera diagnostic system mainly includes the thermograph with uncooled microbolometer Focal Plane Array detector, Zinc Selenide window, Firewire Fiber Repeaters, 50 m long fibers, magnetic shielding box and data acquisition card. The diagnostic system can provide high spatial resolution, long distance control and real-time data acquisition. Based on the surface temperature measured by the infrared camera diagnostic system and the knowledge of the copper thermal properties, the heat flux can be derived by heat conduct model. The infrared camera diagnostic system and preliminary results are presented in details. (authors)

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. Development of a tokamak plasma optimized for stability and confinement

    International Nuclear Information System (INIS)

    Politzer, P.A.

    1995-02-01

    Design of an economically attractive tokamak fusion reactor depends on producing steady-state plasma operation with simultaneous high energy density (β) and high energy confinement (τ E ); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H≡ τ E /τ ITER-89P = 4) and high normalized β (β N ≡ β/(I/aB) = 6%-m-T/MA) have been obtained. For the present, these conditions have been produced separately and in transient discharges. The DIII-D advanced tokamak development program is directed toward developing an understanding of the characteristics which lead to high stability and confinement, and to use that understanding to demonstrate stationary, high performance operation through active control of the plasma shape and profiles. The authors have identified some of the features of the operating modes in DIII-D that contribute to better performance. These are control of the plasma shape, control of both bulk plasma rotation and shear in the rotation and Er profiles, and particularly control of the toroidal current profiles. In order to guide their future experiments, they are developing optimized scenarios based on their anticipated plasma control capabilities, particularly using fast wave current drive (on-axis) and electron cyclotron current drive (off-axis). The most highly developed model is the second-stable core VH-mode, which has a reversed magnetic shear safety factor profile [q(O) = 3.9, q min = 2.6, and q 95 = 6]. This model plasma uses profiles which the authors expect to be realizable. At β N ≥ 6, it is stable to n=l kink modes and ideal ballooning modes, and is expected to reach H ≥ 3 with VH-mode-like confinement

  11. Development in Diagnostics Application to Control Advanced Tokamak Plasma

    International Nuclear Information System (INIS)

    Koide, Y.

    2008-01-01

    For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

  12. Time-dependent analysis of the resistivity of post-disruption tokamak plasmas

    International Nuclear Information System (INIS)

    Bakhtiari, M.; Whyte, D. G.

    2006-01-01

    The effect of neutrals on plasma resistivity due to electron-neutral collisions is studied with respect to its effect on tokamak disruptions. The resistivity of the tokamak plasma after the thermal quench is critical in determining the current quench rate, the plasma temperature, and runaway electron generation in tokamaks through the electric field, all features which are important for mitigating the damaging effect of disruptions. It is shown that the plasma resistivity during tokamak disruptions is a time-dependent parameter which may vary with disruption time scales due to the increasing fraction of neutrals. However the effect of neutrals on resistivity is found to be small for the expected neutral fraction, mostly due to power balance considerations between radiation and Ohmic heating in the plasma

  13. Electron cyclotron heating (ECH) of tokamak plasmas

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi

    1990-01-01

    Electron cyclotron heating (ECH) is one of the intense methods of plasma heating, and which utilizes the collisionless electron-cyclotron-resonance-interaction between the launched electromagnetic waves (called electron cyclotron waves) and electrons which are one of the constituents of the high temperature plasmas. Another constituent, namely the ions which are subject to nuclear fusion, are heated indirectly but strongly and instantly (in about 0.1 s) by the collisions with the ECH-heated electrons in the fusion plasmas. The recent progress on the development of high-power and high-frequency millimeter-wave-source enabled the ECH experiments in the middle size tokamaks such as JFT-2M (Japan), Doublet III (USA), T-10 (USSR) etc., and ECH has been demonstrated to be the sure and intense plasma heating method. The ECH attracts much attention for its remarkable capabilities; to produce plasmas (pre-ionization), to heat plasmas, to drive plasma current for the plasma confinement, and recently especially by the localization and the spatial controllability of its heating zone, which is beneficial for the fine controls of the profiles of plasma parameters (temperature, current density etc.), for the control of the magnetohydrodynamic instabilities, or for the optimization/improvement of the plasma confinement characteristics. Here, the present status of the ECH studies on tokamak plasmas are reviewed. (author)

  14. Study of fast wave current drive in a KT-2 tokamak plasma

    International Nuclear Information System (INIS)

    Hong, B.G.; Hamamatsu, Kiyotaka

    1996-02-01

    Global analysis of fast wave current drive in a KT-2 tokamak plasma is performed by using the code, TASKW1, developed by JAERI and Okayama University (Dr. Fukuyama), which solves the kinetic wave equation in a one dimensional slab geometry. A phase-shifted antenna array is used to inject toroidal momentum to electrons. To find guidelines of optimum antenna design for efficient current drive, accessibility conditions are derived. The dependence of the current drive efficiency on launching conditions such as the total number of antennas, phase and spacing is investigated for two cases of wave frequency; f=30 MHz ( cH ) and f=225 MHz (=5f cH ). (author)

  15. Mathematical modeling plasma transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)

    1997-01-01

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

  16. Mathematical modeling plasma transport in tokamaks

    International Nuclear Information System (INIS)

    Quiang, Ji

    1995-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

  17. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    Pitcher, C.S.

    1987-11-01

    The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D 2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict

  18. Relativistic runaway electrons in tokamak plasmas

    International Nuclear Information System (INIS)

    Jaspers, R.E.

    1995-01-01

    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)

  19. Design of Tokamak plasma with high Tc superconducting coils

    International Nuclear Information System (INIS)

    Uchimoto, T.; Miya, K.; Yoshida, Y.; Yamada, T.

    1999-01-01

    This paper presents a design of tokamak plasma in light of how the small ignited tokamak is possible with use of the HTSC coils as plasma stabilizer. The same data base and formulas as ITER are here used and any innovative technology other than the HTSC stabilizing coils is not assumed. (author)

  20. Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment

    Science.gov (United States)

    Lucia, Matthew James

    The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (dLi ˜ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2 O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H 2 within minutes. For impurity sequestration, LTX plasma performance

  1. Neutral-beam deposition in large, finite-beta noncircular tokamak plasmas

    International Nuclear Information System (INIS)

    Wieland, R.M.; Houlberg, W.A.

    1982-02-01

    A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross section and noncircular shifted plasma cross sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is benchmarked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radius a = 120 cm and elongation b/a = 1.6. Deposition profiles are compared for deuterium beam energies of 120 to 150 keV, central plasma densities of 8 x 10 13 - 2 x 10 14 cm -3 , and beam orientation ranging from perpendicular to tangential to the inside wall

  2. Control of plasma position in the CASTOR tokamak

    International Nuclear Information System (INIS)

    Valovic, M.

    1988-11-01

    A simple servo-system designed for plasma position control in the CASTOR tokamak is described. Both radial and vertical plasma displacements were minimized using two servo-loops consisting of detection coils, a conventional electric controller and an amplifier operated as an unipolar voltage-controlled current source. To ensure the optimum conditions in the start-up phase of the discharge, currents in the servo-systems were externally preprogrammed. The prescribed plasma position was maintained with the accuracy of 3 mm. The feedback control improves plasma parameters, e.g. it removes the positional disruption at the end of the tokamak discharge. (J.U.). 4 figs., 3 refs

  3. A study on the fusion reactor - Development of x-ray spectrometer for diagnosis of tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hong Young; Choi, Duk In; Seo, Sung Hun; Kwon, Gi Chung; Jun, Sang Jin; Heo, Sung Hoi; Lee, Chan Hui [Korea Advanced Institute of Science and Technolgoy, Taejon (Korea, Republic of)

    1996-09-01

    This report of research is on the development of X-ray Photo-Electron Spectrometer (PES) for diagnosis of tokamak plasma. The spectrometer utilizes the fact that the energy of photo-electron is given by the difference between the energy of X-ray and the binding energy of materials. In the research of this year, we constructed two spectrometers; one is operated in KAIST tokamak and the other in KT1 tokamak. In addition, we reviewed the characteristics of the x-ray filter, the photo-electric effect of carbon foils and the detection efficiency of MCP and x-ray radiation of plasma. We measured the x-ray radiation in tokamak and diagnosed the qualitative plasma parameters from the analysis of data. The major interesting plasma parameters, which we can diagnose with the spectrometer, are the electron temperature, Z{sub eff}, the spatial distribution of x-ray radiation and etc. 27 refs., 2 tabs., 20 figs. (author)

  4. ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

    International Nuclear Information System (INIS)

    HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

    2002-01-01

    A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

  5. Design of plasma facing components for the SST-1 tokamak

    International Nuclear Information System (INIS)

    Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

    2000-01-01

    Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

  6. Magnetic diagnostic plasma position in the TCA/BR tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu.K.; Nascimento, I.C.

    1996-01-01

    The cross-section of the plasma column is TCA/BR has a nearly circular plasma shape. This allows implementation of simplified methods of magnetic diagnostics. Although these methods were in may tokamaks and are well described, their accuracies are not clearly defined because the very simplified theoretical model of plasma equilibrium on which they are based differs from the real conditions in tokamaks like TCA/BR. In this paper we present the methods of plasma position diagnostics in TCA/BR from external magnetic measurements with an error analysis. (author). 4 refs., 3 figs

  7. Density peaking in the JFT-2M tokamak plasma with counter neutral beam injection

    International Nuclear Information System (INIS)

    Ida, K.; Itoh, S.; Itoh, K.

    1991-05-01

    A significant particle pinch and reduction of the effective thermal diffusivity are observed after switching the neutral beam direction from co- to counter- injection in the JFT-2M tokamak. A time delay in the occurrence of density peaking to that of plasma rotation is found. This shows that the particle pinch is related to the profile of the electric field as determined by the plasma rotation profile. The measured particle flux shows qualitative agreement with the theoretically-predicted inward pinch. (author)

  8. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

    The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

  9. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  10. The method of inversion of magnetic island two-dimensional structure by magnetic probes and its application on HL-2A tokamak

    International Nuclear Information System (INIS)

    Sun Tengfei; Liu Yi; Ji Xiaoquan; Xu Yuan; Feng Beibin

    2011-01-01

    The new method that reconstructs the polar two-dimensional structure of the magnetic island using magnetic pickup coils data is introduced on HL-2A tokamak and dynamic analysis method that set up based on it for tearing mode is also introduced. In this experiment, the perturbation current which is the source of the perturbation magnetic field can be determined using the data measured by magnetic probes. Superimposing the perturbation flux and equilibrium flux reconnected by EFIT, the structure and the width of the magnetic islands can be obtained. Then two-dimensional structure maps are set up chronologically and recorded in turn. After that these maps are revealed in turn and magnetic island can be analyzed dynamically. This method is applied to analyzing tearing mode. The conclusion that magnetic island rotating direction is in accordance with electronic diamagnetic drift direction is reached. The relationship between the magnetic island width and the magnetic perturbation field is proved and the suppression of magnetic island by ECRH is also verified.It shows the immediacy of the method of inversion of magnetic island structure by magnetic probes and it is very useful for watching and controlling MHD instability. (authors)

  11. Magnetic evaluation of hydrogen pressures changes on MHD fluctuations in IR-T1 tokamak plasma

    Science.gov (United States)

    Alipour, Ramin; Ghanbari, Mohamad R.

    2018-04-01

    Identification of tokamak plasma parameters and investigation on the effects of each parameter on the plasma characteristics is important for the better understanding of magnetohydrodynamic (MHD) activities in the tokamak plasma. The effect of different hydrogen pressures of 1.9, 2.5 and 2.9 Torr on MHD fluctuations of the IR-T1 tokamak plasma was investigated by using of 12 Mirnov coils, singular value decomposition and wavelet analysis. The parameters such as plasma current, loop voltage, power spectrum density, energy percent of poloidal modes, dominant spatial structures and temporal structures of poloidal modes at different plasma pressures are plotted. The results indicate that the MHD activities at the pressure of 2.5 Torr are less than them at other pressures. It also has been shown that in the stable area of plasma and at the pressure of 2.5 Torr, the magnetic force and the force of plasma pressure are in balance with each other and the MHD activities are at their lowest level.

  12. On steady poloidal and toroidal flows in tokamak plasmas

    International Nuclear Information System (INIS)

    McClements, K. G.; Hole, M. J.

    2010-01-01

    The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B θ /B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B θ /B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.

  13. Plasma surface engineering in first wall of tokamak

    International Nuclear Information System (INIS)

    Liu Xiang; Xu Zengyu; Zhang Fu; Zhang Nianman

    2001-01-01

    The boronization, siliconization and lithium coating of the inner wall of HL-1M are introduced, the hydrogen recycling and the influence to impurities controlled and core radiation energy loss are discussed. Experiments prove that these wall treatments are very useful for the plasma confinement, a 4 s reproducible long pulse discharge is obtained for siliconized wall, but the plasma pulse length only achieves 2.1 s and its reproducibility is very poor for boronized wall. Lithium coating is the best method of the wall treatments for lowering hydrogen recycling and decreasing the impurities level. For the applications of HL-2A and the future fusion device, a series of B, Ti, Si-doped graphite and B 4 C-C/C composites have been developed, some experimental results about chemical sputtering, tritium retention and recycling, as well as high heat loads are reviewed. Meanwhile, SiC, TiC and B 4 C coating, and B 4 C-C, SiC-C, B 4 C-Cu, Mo-Cu and W-Cu functionally graded materials are also introduced

  14. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)

  15. A flexible software design to determine the plasma boundary in Damavand tokamak

    International Nuclear Information System (INIS)

    Ghadiri, R.; Esteki, M.H.; Sadeghi, Y.

    2014-01-01

    A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code “The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)” was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C language. (author)

  16. Kinetic modelling of runaway electron avalanches in tokamak plasmas.

    Czech Academy of Sciences Publication Activity Database

    Nilsson, E.; Decker, J.; Peysson, Y.; Granetz, R.S.; Saint-Laurent, F.; Vlainic, Milos

    2015-01-01

    Roč. 57, č. 9 (2015), č. článku 095006. ISSN 0741-3335 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : plasma physics * runaway electrons * knock-on collisions * tokamak * Fokker-Planck * runaway avalanches Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.404, year: 2015

  17. Formation and sustainment of a low aspect ratio tokamak by a series of plasma injections

    International Nuclear Information System (INIS)

    Shimamura, Shin; Taniguchi, Makoto; Takahashi, Tsutomu; Nogi, Yasuyuki

    1995-01-01

    A low aspect ratio tokamak plasma was generated and sustained by injecting a series of plasmas from a magnetized coaxial gun into a flux conserver with toroidal field. The magnetized coaxial gun was supplied by an oscillating current with a d.c. component. The first few current pulses injected plasma and helicity into the flux conserver. This pulse helicity injection method worked effectively to maintain the low aspect ratio tokamak. 8 refs., 5 figs

  18. Comparison of bootstrap current and plasma conductivity models applied in a self-consistent equilibrium calculation for Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

    Different bootstrap current formulations are implemented in a self-consistent equilibrium calculation obtained from a direct variational technique in fixed boundary tokamak plasmas. The total plasma current profile is supposed to have contributions of the diamagnetic, Pfirsch-Schlueter, and the neoclassical Ohmic and bootstrap currents. The Ohmic component is calculated in terms of the neoclassical conductivity, compared here among different expressions, and the loop voltage determined consistently in order to give the prescribed value of the total plasma current. A comparison among several bootstrap current models for different viscosity coefficient calculations and distinct forms for the Coulomb collision operator is performed for a variety of plasma parameters of the small aspect ratio tokamak ETE (Experimento Tokamak Esferico) at the Associated Plasma Laboratory of INPE, in Brazil. We have performed this comparison for the ETE tokamak so that the differences among all the models reported here, mainly regarding plasma collisionality, can be better illustrated. The dependence of the bootstrap current ratio upon some plasma parameters in the frame of the self-consistent calculation is also analysed. We emphasize in this paper what we call the Hirshman-Sigmar/Shaing model, valid for all collisionality regimes and aspect ratios, and a fitted formulation proposed by Sauter, which has the same range of validity but is faster to compute than the previous one. The advantages or possible limitations of all these different formulations for the bootstrap current estimate are analysed throughout this work. (author)

  19. Plasma Sprayed Tungsten-based Coatings and their Usage in Edge Plasma Region of Tokamaks

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Dufková, Edita; Piffl, Vojtěch; Peřina, Vratislav

    2006-01-01

    Roč. 51, č. 2 (2006), s. 179-191 ISSN 0001-7043 Grant - others:Evropská unie EFDA Task TW-5-TVM-PSW (EU – Euratom) Institutional research plan: CEZ:AV0Z20430508; CEZ:AV0Z10480505 Keywords : plasma sprayed coatings * fusion * plasma facing components * tungsten * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics

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

    International Nuclear Information System (INIS)

    Martin, P.; Castro, E.

    2007-01-01

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

  1. IR and FIR laser polarimetry as a diagnostic tool in high-. beta. and Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, D; Machida, M; Scalabrin, A

    1986-03-01

    The change of the polarization state of an electromagnetic wave (EMW) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the EMW allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-pinch plasmas (lambda approx. 10..mu..m). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-pinch/UNICAMP, using lasers developed at UNICAMP.

  2. About the Toroidal Magnetic Field of a Tokamak Burning Plasma Experiment with Superconducting Coils

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2002-01-01

    In tokamaks, the strong dependence on the toroidal magnetic field of both plasma pressure and energy confinement is what makes possible the construction of small and relatively inexpensive burning plasma experiments using high-field resistive coils. On the other hand, the toroidal magnetic field of tokamaks using superconducting coils is limited by the critical field of superconductivity. In this article, we examine the relative merit of raising the magnetic field of a tokamak plasma by increasing its aspect ratio at a constant value of the peak field in the toroidal magnet. Taking ITER-FEAT as an example, we find that it is possible to reach thermonuclear ignition using an aspect ratio of approximately 4.5 and a toroidal magnetic field of 7.3 T. Under these conditions, fusion power density and neutron wall loading are the same as in ITER [International Thermonuclear Experimental Reactor], but the normalized plasma beta is substantially smaller. Furthermore, such a tokamak would be able to reach an energy gain of approximately 15 even with the deterioration in plasma confinement that is known to occur near the density limit where ITER is forced to operate

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

    International Nuclear Information System (INIS)

    Evans, T.E.

    1984-09-01

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

  4. Plasma features and alpha particle transport in low-aspect ratio tokamak reactor

    International Nuclear Information System (INIS)

    Xu Qiang; Wang Shaojie

    1997-06-01

    The results of the experiment and theory from low-aspect ratio tokamak devices have proved that the MHD stability will be improved. Based on present plasma physics and extrapolation to reduced aspect ratio, the feature of physics of low-aspect ratio tokamak reactor is discussed primarily. Alpha particle confinement and loss in the self-justified low-aspect ratio tokamak reactor parameters and the effect of alpha particle confinement and loss for different aspect ratio are calculated. The results provide a reference for the feasible research of compact tokamak reactor. (9 refs., 2 figs., 3 tabs.)

  5. Control of plasma poloidal shape and position in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

  6. Practical aspects of a 2-D edge-plasma model

    International Nuclear Information System (INIS)

    Rensink, M.E.; Hill, D.N.; Porter, G.D.; Braams, B.J.; Princeton Univ., NJ

    1989-07-01

    The poloidal divertor configuration is considered the most promising solution to the particle and energy exhaust problem for a tokamak reactor. The scrape-off layer plasma surrounding the core and the high-recycling plasma near the divertor plates can be modelled by fluid equations for particle, momentum and energy transport. A numerical code (B2) based on a two-dimensional multi-fluid model has been developed for the study of edge plasmas in tokamaks. In this report we identify some key features of this model as applied to the DIII-D tokamak. 2 refs., 1 fig

  7. Detection of tokamak plasma positrons using annihilation photons

    Energy Technology Data Exchange (ETDEWEB)

    Guanying, Yu; Liu, Jian; Xie, Jinlin [University of Science and Technology, Hefei, Anhui, 230027 (China); Li, Jiangang, E-mail: j_li@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2017-05-15

    Highlights: • A design for detection of tokamak plasma positrons is given. • Identify the main obstacle toward experimental confirmation of fusion plasma positrons. • Signal to noise ratio in a plasma disruption is estimated. • Unique potential applications of fusion plasma positrons are discussed. - Abstract: A massive amount of positrons (plasma positrons), produced by the collision between runaway electrons and nuclei during fusion plasma disruption, was first predicted theoretically in 2003. To help confirm this prediction, we report here the design of an experimental system to detect tokamak plasma positrons. Because a substantial amount of positrons (material positrons) are produced when runaway electrons impact plasma-facing materials, we proposed maximizing the ratio of plasma to material positrons by inserting a thin carbon target at the plasma edge as a plasma positron bombing target and producing a plasma disruption scenario triggered by massive gas injection. Meanwhile, the coincidence detection of positron annihilation photons was used to filter out the noise of annihilation photons from locations other than the carbon target and that of bremsstrahlung photons near 511 keV. According to our simulation, the overall signal-to-noise ratio should be more than 10:1.

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

    Czech Academy of Sciences Publication Activity Database

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

    2008-01-01

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

  9. Application studies of spherical tokamak plasma merging

    International Nuclear Information System (INIS)

    Ono, Yasushi; Inomoto, Michiaki

    2012-01-01

    The experiment of plasma merging and heating has long history in compact torus studies since Wells. The study of spherical tokamak (ST), starting from TS-3 plasma merging experiment of Tokyo University in the late 1980s, is followed by START of Culham laboratory in the 1900s, TS-4 and UTST of Tokyo University and MAST of Culham laboratory in the 2000s, and last year by VEST of Soul University. ST has the following advantages: 1) plasma heating by magnetic reconnection at a MW-GW level, 2) rapid start-up of high beta plasma, 3) current drive/flux multiplication and distribution control of ST plasma, 4) fueling and helium-ash exhaust. In the present article, we emphasize that magnetic reconnection and plasma merging phenomena are important in ST plasma study as well as in plasma physics. (author)

  10. Application of mineral insulated cable (MIC) in Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Luo Tianyong; Jiang Jiaming; Cen Yishun

    2014-01-01

    To avoid the instability of plasma and achieve some experimental tasks in Tokamak fusion reactor, many in-vessel coils are designed such as the coils to mitigate the effect of Edge Localized Modes (ELMs coils) and the coils to provide vertical stabilization (VS coils). The in-vessel location presents special challenges in terms of nuclear radiation and temperature, and requires the use of mineral-insulated conductors. The in-vessel coils in ITER are designed to be Mineral-insulated Cable (MIC) with three-layer structures. The inner is hollow-core tube made by OFHC or CuCrZr, the middle is the insulation layer made by Mgo and the outer is the jacket by SS316L or Inconel 718. To control the effect of Edge Localized Modes and vertical instability of plasma, the MIC in-vessel coils shall be used in HL-2M. More details about the application of MIC in Tokamak fusion reactor will be shown in this report. (authors)

  11. Models for Predicting Boundary Conditions in L-Mode Tokamak Plasma

    International Nuclear Information System (INIS)

    Siriwitpreecha, A.; Onjun, T.; Suwanna, S.; Poolyarat, N.; Picha, R.

    2009-07-01

    Full text: The models for predicting temperature and density of ions and electrons at boundary conditions in L-mode tokamak plasma are developed using an empirical approach and optimized against the experimental data obtained from the latest public version of the International Pedestal Database (version 3.2). It is assumed that the temperature and density at boundary of L-mode plasma are functions of engineering parameters such as plasma current, toroidal magnetic field, total heating power, line averaged density, hydrogenic particle mass (A H ), major radius, minor radius, and elongation at the separatrix. Multiple regression analysis is carried out for these parameters with 86 data points in L-mode from Aug (61) and JT60U (25). The RMSE of temperature and density at boundary of L-mode plasma are found to be 24.41% and 18.81%, respectively. These boundary models are implemented in BALDUR code, which will be used to simulate the L-mode plasma in the tokamak

  12. 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.

  13. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    Caldas, I.L.; Vannucci, A.

    1985-01-01

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt

  14. Plasma current profile during current reversal in a tokamak

    International Nuclear Information System (INIS)

    Huang Jianguo; Yang Xuanzong; Zheng Shaobai; Feng Chunhua; Zhang Houxian; Wang Long

    1999-01-01

    Alternating current operation with one full cycle and a current level of 2.5 kA have been achieved in the CT-6B tokamak. The poloidal magnetic field in the plasma is measured with two internal magnetic probes in repeated discharges. The current distribution is reconstructed with an inversion algorithm. The inverse current first appears on the weak field side. The existence of magnetic surfaces and rotational transform provide particle confinement in the current reversal phase

  15. Scrape-off layer tokamak plasma turbulence

    Science.gov (United States)

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

    2012-05-01

    Two-dimensional (2D) interchange turbulence in the scrape-off layer of tokamak plasmas and their subsequent contribution to anomalous plasma transport has been studied in recent years using electron continuity, current balance, and electron energy equations. In this paper, numerically it is demonstrated that the inclusion of ion energy equation in the simulation changes the nature of plasma turbulence. Finite ion temperature reduces floating potential by about 15% compared with the cold ion temperature approximation and also reduces the radial electric field. Rotation of plasma blobs at an angular velocity about 1.5×105 rad/s has been observed. It is found that blob rotation keeps plasma blob charge separation at an angular position with respect to the vertical direction that gives a generation of radial electric field. Plasma blobs with high electron temperature gradients can align the charge separation almost in the radial direction. Influence of high ion temperature and its gradient has been presented.

  16. Plasma startup patterns in tokamak reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Tone, Tatsuzo.

    1983-01-01

    Plasma startup patterns are studied from the viewpoint of net power loss represented by the total power loss less the α-particle heating power. The existence is shown of a critical temperature of plasma at which the net power loss becomes independent of plasma density. Observations are made which indicate that the net power loss decreases with lowering plasma density in the range below the critical temperature and vice versa, whether governed by empirical or trapped-ion scaling laws. A startup pattern is presented which minimizes the net power loss during startup, and which prescribes that: (1) The plasma density should be kept as low as possible until the plasma is heated up to the critical temperature; (2) thereafter, the plasma density should be increased to its steady state value while retaining the critical temperature; and (3) finally, with the density kept constant, the temperature should be further raised to its steady state value. The net power loss at critical temperature represents the lower limit of heating power required to bring the plasma to steady state in tokamak reactors. (author)

  17. Study of impurity transport in HL-2A ECRH L-mode plasmas with radially different ECRH power depositions

    Science.gov (United States)

    Cui, Z. Y.; Zhang, K.; Morita, S.; Ji, X. Q.; Ding, X. T.; Xu, Y.; Sun, P.; Gao, J. M.; Dong, C. F.; Zheng, D. L.; Li, Y. G.; Jiang, M.; Li, D.; Zhong, W. L.; Liu, Yi; Dong, Y. B.; Song, S. D.; Yu, L. M.; Shi, Z. B.; Fu, B. Z.; Lu, P.; Huang, M.; Yuan, B. S.; Yang, Q. W.; Duan, X. R.

    2018-05-01

    In HL-2A, an inverse sawtooth oscillation is observed with a long-lasting m/n  =  1/1 mode during ECRH phase with power deposition inside sawtooth inversion radius (inner-deposited ECRH), while a normal sawtooth instead appears when the ECRH power is deposited outside sawtooth inversion radius (outer-deposited ECRH). Aluminum is then injected as a trace impurity with laser blow-off (LBO) method into the inner- and outer-deposited ECRH phases of HL-2A discharges to investigate the effect of ECRH on impurity transport. Temporal behavior of soft x-ray (SXR) array signals is analyzed with a 1D impurity transport code, and radial structures of impurity transport coefficients are obtained. The result shows that the radial transport of Al ions is strongly enhanced during the inner-deposited ECRH phase. In particular, an outward convection velocity is developed with positive values of 0  ⩽  V(ρ)  ⩽  3.8 m s-1 in ρ  ⩽  0.5, while the convection velocity is inward in ρ  ⩾  0.6. In the outer-deposited ECRH discharge, on the other hand, the convection velocity takes a big negative value in ρ  ⩽  0.4 and close to zero at ρ ~ 0.6. In ohmic discharges, an inward V(ρ) always appears in the whole plasma radii and gradually increases toward the plasma edge (-3.2 m s-1 at ρ  =  1). The simulation result also indicates that centrally-peaked Al ion density profiles presented in the outer-deposited ECRH discharge can be flattened by the inner-deposited ECRH. Modification of impurity transport is discussed in the presence of long-lasting m/n  =  1/1 MHD mode.

  18. Plasma residual poloidal rotation in TCABR tokamak

    International Nuclear Information System (INIS)

    Severo, J.H.F.; Nascimento, I.C.; Tsypin, V.S.; Galvao, R.M.O.

    2003-01-01

    This paper reports the first measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak for a collisional plasma (Pfirsch-Schluter regime), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5±1.0)·10 3 m/s at r ∼ 2/3a, (a - limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) · 10 3 m/s for the plasma core to a co-current value of (2.0 ± 1.0) · 10 3 m/s near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner. (author)

  19. Turbulence in tokamak plasmas. Effect of a radial electric field shear

    International Nuclear Information System (INIS)

    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

  20. Study of intelligent system for control of the tokamak-ETE plasma positioning

    International Nuclear Information System (INIS)

    Barbosa, Luis Filipe de Faria Pereira Wiltgen

    2003-01-01

    The development of an intelligent neural control system of the neural type, capable to perform real time control of the plasma displacement in the experiment tokamak spheric - ETE (spherical tokamak experiment ) is presented. The ETE machine is in operation since Nov 2000, in the LAP - Plasma Associated Laboratory of the Brazilian Institute on Spatial Research (INPE) in Sao Jose dos Campos, S P, Brazil. The experiment is dedicated to study the magnetic confinement of a fusion plasma in a configuration favorable for the construction of future reactors. Nuclear fusion constitutes a renewable energy source with low environmental impact, which uses atomic energy in pacific applications for the sustainable development of humanity. One of the important questions for the attainment of fusion relates to the stability of the plasma and control of its position during the reactor operation. Therefore, the development of systems to control the plasma in tokamaks constitutes a necessary technological advance for the feasibility of nuclear fusion. In particular, the research carried out in this thesis concerns the proposal of a system to control the vertical displacement of the plasma in the ETE tokamak, aiming to obtain steady pulses in this machine. A Magnetic Levitation system (Mag Lev) was developed as part of this work, allowing to study the nonlinear behavior of a device that, from the aspect of position control, is similar (analogous) to the plasma in the ETE tokamak, This magnetic levitation system was designed, mathematically modeled and built in order to test both classical and intelligent type controllers. The results of this comparison are very promising for the use of intelligent controllers in the ETE tokamak as well as other control applications. (author)

  1. Experimental observations related to the thermodynamic properties of tokamak plasmas

    International Nuclear Information System (INIS)

    Sozzi, C.; Minardi, E.; Lazzaro, E.; Cirant, S.; Mantica, P.; Esposito, B.; Marinucci, M.; Romanelli, M.; Imbeaux, F.

    2005-01-01

    The coarse-grained tokamak plasma description derived from the magnetic entropy concept presents appealing features as it involves a simple mathematics and it identifies a limited set of characteristic parameters of the macroscopic equilibrium. In this paper a comprehensive review of the work done in order to check the reliability of the Stationary Magnetic Entropy predictions against experimental data collected from different tokamaks, plasma regimes and heating methods is reported. (author)

  2. Collisional boundary layer analysis for neoclassical toroidal plasma viscosity in tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    2008-01-01

    Roč. 15, č. 8 (2008), 082506-1-7 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2969434

  3. Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma

    International Nuclear Information System (INIS)

    Selig, G.

    2012-12-01

    In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)

  4. Energy confinement of tokamak plasma with consideration of bootstrap current effect

    International Nuclear Information System (INIS)

    Yuan Ying; Gao Qingdi

    1992-01-01

    Based on the η i -mode induced anomalous transport model of Lee et al., the energy confinement of tokamak plasmas with auxiliary heating is investigated with consideration of bootstrap current effect. The results indicate that energy confinement time increases with plasma current and tokamak major radius, and decreases with heating power, toroidal field and minor radius. This is in reasonable agreement with the Kaye-Goldston empirical scaling law. Bootstrap current always leads to an improvement of energy confinement and the contraction of inversion radius. When γ, the ratio between bootstrap current and total plasma current, is small, the part of energy confinement time contributed from bootstrap current will be about γ/2

  5. Plasma diagnostics for the compact ignition tokamak

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  6. 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.

  7. The effects of plasma deformability on the feedback stabilization of axisymmetric modes in tokamak plasmas

    International Nuclear Information System (INIS)

    Ward, D.J.; Jardin, S.C.

    1991-09-01

    The effects of plasma deformability on the feedback stabilization of axisymmetric modes of tokamak plasmas are studied. It is seen that plasmas with strongly shaped cross sections have unstable motion different from a rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the eigenfunction in a way that reduces the stabilizing eddy currents in these conductors. Passive feedback results using several equilibria of varying shape are presented. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops which are used to determine plasma vertical position for the active feedback system. The variations of these non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations. Active feedback results are presented for the PBX-M tokamak configuration. (author) 19 figs., 2 tabs., 30 refs

  8. The importance of the toroidal magnetic field for the feasibility of a tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2000-01-01

    The next step in the demonstration of the scientific feasibility of a tokamak fusion reactor is a DT burning plasma experiment for the study and control of self-heated plasmas. In this paper, the authors examine the role of the toroidal magnetic field on the confinement of a tokamak plasma in the ELMy H-mode regime--the operational regime foreseen for ITER

  9. Impurity screening of scrape-off plasma in a tokamak

    International Nuclear Information System (INIS)

    Kishimoto, Hiroshi; Tani, Keiji; Nakamura, Hiroo

    1981-11-01

    Impurity screening effect of a scrape-off layer has been studied in a tokamak, based on a simple model of wall-released impurity behavior. Wall-sputtered impurities are stopped effectively by the scrape-off plasma for a medium-Z or high-Z wall system while major part of impurities enters the main plasma in a low-Z wall system. The screening becomes inefficient with increase of scrape-off plasma temperature. Successive multiplication of recycling impurities in the scrape-off layer is large for a high-Z wall and is enhanced by a rise of scrape-off plasma temperature. The stability of plasma-wall interaction is determined by a multiplication factor of recycling impurities. (author)

  10. Radial electric field at the plasma edge on the FT-2 Tokamak in regimes with large gradients

    International Nuclear Information System (INIS)

    Lashkul, S.; Popov, A.

    2001-01-01

    The transport barrier formation is widely believed to be the fundamental element of transition into improved confinement regimes (H-mode). Experiments on many tokamaks demonstrate that transport barrier formation is connected with the suppression of turbulent transport by shear of E x B drift. Therefore, the calculation of radial electric field is of great importance. Our work is devoted to progress the neoclassical theory by taking into account electron viscosity and non-linear effects (ion inertia), presented results being valuable for interpretation transition into H-mode at the plasma edge in small tokamaks. Calculations of the electric field profile for FT-2 tokamak (a=8cm, R 0 =55cm, Ioffe Institute, Russia) according found expressions are in the good agreement with experimental results obtained. (orig.)

  11. 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.

  12. Extremely shaped plasmas to improve the Tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Piras, F.

    2011-04-15

    presented, and recent attempts to create a doublet plasma are reported. Since the magnetic field reconstruction page at the breakdown time is important to better diagnose these plasmas, the entire magnetic system of TCV has been calibrated with an original technique, also described in the manuscript. The last part of this thesis is devoted to the snowflake divertor configuration. This innovative plasma shape has been proposed and theoretically studied by Dr. D.D. Ryutov from the Lawrence Livermore National Laboratory. In Ryutov’s articles, this configuration was proposed to alleviate the problems of the plasma-wall interaction and possibly affect the plasma edge stability. The TCV tokamak was the first to report the creation and control of a snowflake configuration, and the candidate was the principal investigator of this work. These results are accordingly discussed in this thesis. Details are provided in particular on the strategy used to establish the configuration. An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a snowflake. This regime exhibits 2 to 3 times lower ELM frequency but only a 20%-30% increase in normalized ELM energy (ΔW{sub ELM}/W{sub P} ) compared to an identically-shaped, conventional, single-null, diverted H-mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. Finally, the capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally and is also reported in this thesis. (author)

  13. Extremely shaped plasmas to improve the Tokamak concept

    International Nuclear Information System (INIS)

    Piras, F.

    2011-04-01

    presented, and recent attempts to create a doublet plasma are reported. Since the magnetic field reconstruction page at the breakdown time is important to better diagnose these plasmas, the entire magnetic system of TCV has been calibrated with an original technique, also described in the manuscript. The last part of this thesis is devoted to the snowflake divertor configuration. This innovative plasma shape has been proposed and theoretically studied by Dr. D.D. Ryutov from the Lawrence Livermore National Laboratory. In Ryutov’s articles, this configuration was proposed to alleviate the problems of the plasma-wall interaction and possibly affect the plasma edge stability. The TCV tokamak was the first to report the creation and control of a snowflake configuration, and the candidate was the principal investigator of this work. These results are accordingly discussed in this thesis. Details are provided in particular on the strategy used to establish the configuration. An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a snowflake. This regime exhibits 2 to 3 times lower ELM frequency but only a 20%-30% increase in normalized ELM energy (ΔW ELM /W P ) compared to an identically-shaped, conventional, single-null, diverted H-mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. Finally, the capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally and is also reported in this thesis. (author)

  14. Sensitivity of transient synchrotron radiation to tokamak plasma parameters

    International Nuclear Information System (INIS)

    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

  15. Analysis of plasma flow in a scrape-off layer in a tokamak

    International Nuclear Information System (INIS)

    Petrov, V.G.

    1988-01-01

    Plasma drift on the periphery of a tokamak in a magnetic tube, on sides of which coins are placed, is considered. Convection caused by toroidal particle drift is taken into account. Distribution of plasma parameters in such a tube is found. Transition from the total poloidal diaphragm to a sectioned one is traced

  16. Transport in the tokamak plasma edge

    International Nuclear Information System (INIS)

    Vold, E.L.

    1989-01-01

    Experimental observations characterize the edge plasma or boundary layer in magnetically confined plasmas as a region of great complexity. Evidence suggests the edge physics plays a key role in plasma confinement although the mechanism remains unresolved. This study focuses on issues in two areas: observed poloidal asymmetries in the Scrape Off Layer (SOL) edge plasma and the physical nature of the plasma-neutral recycling. A computational model solves the coupled two dimensional partial differential equations governing the plasma fluid density, parallel and radial velocities, electron and ion temperatures and neutral density under assumptions of toroidal symmetry, ambipolarity, anomalous diffusive radial flux, and neutral-ion thermal equilibrium. Drift flow and plasma potential are calculated as dependent quantities. Computational results are compared to experimental data for the CCT and TEXTOR:ALT-II tokamak limiter cases. Comparisons show drift flux is a major component of the poloidal flow in the SOL along the tangency/separatrix. Plasma-neutral recycling is characterized in several tokamak divertors, including the C-MOD device using magnetic flux surface coordinates. Recycling is characterized by time constant, τ rc , on the order of tens of milliseconds. Heat flux transients from the core into the edge on shorter time scales significantly increase the plasma temperatures at the target and may increase sputtering. Recycling conditions in divertors vary considerably depending on recycled flux to the core. The high density, low temperature solution requires that the neutral mean free path be small compared to the divertor target to x-point distance. The simulations and analysis support H-mode confinement and transition models based on the recycling divertor solution bifurcation

  17. High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

    International Nuclear Information System (INIS)

    Goodall, D.H.J.

    1982-01-01

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)

  18. Study of Globus-M Tokamak Poloidal System and Plasma Position Control

    Science.gov (United States)

    Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

    2017-12-01

    In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

  19. Neoclassical transport of impurtities in tokamak plasmas

    International Nuclear Information System (INIS)

    Hirshman, S.P.; Sigmar, D.J.

    1981-05-01

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

  20. Effect of plasma density profile of tokamak on Kelvin-Helmholtz instability

    International Nuclear Information System (INIS)

    Tang Fulin

    1984-01-01

    The purpose of this paper is to study the effect of radial distribution of plasma density profile of tokamak on Kelvin-Helmholtz instability caused by toroidal rotation. The effect of radial distribution of plasma rotational velocity on stability is also examine for comparison. It is found that within the range of tokamak parameters the only radial distribution of plasma rotational velocity cannot induce Kelvin-Helmholtz instability. On the contrary, when there is a radial distribution of plasma density, i.e. P 01 =P 0 e -tx and V 0 1 = const, plasma becomes unstable, and instability will increase proportionally to the value of t. Meanwhile when the value of t remains constant, the instability growth rate will decrease if P 0 grows or the distance between plasma and wall of container decreases too. It shows that the Kelvin-Helmoltz instability is not only influenced by the steepness of density profile but also by the inertia of plasma in central region, which is helpful for depressing the instability. (author). 5 refs, 4 figs, 2 tabs

  1. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

    Wesson, John; Campbell, D.J.; Connor, J.W.

    1997-01-01

    It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

  2. Simulations of the operational control of a cryogenic plant for a superconducting burning-plasma tokamak

    CERN Document Server

    Mitchell, N

    2001-01-01

    In recent proposals for next generation superconducting tokamaks, such as the ITER project, the nuclear burning plasma is confined by magnetic fields generated from a large set (up to 100 GJ stored energy) of superconducting magnets. These magnets suffer heat loads in operation from thermal and nuclear radiation from the surrounding components and plasma as well as eddy currents and AC losses generated within the magnets, together with the heat conduction through supports and resistive heat generated at the current lead transitions to room temperature. The initial cryoplant for such a tokamak is expected to have a steady state capacity of up to about 85 kW at 4.5 K, comparable to the system installed for LHC at CERN. Experimental tokamaks are expected to operate at least initially in a pulsed mode with 20-30 short plasma pulses and plasma burn periods each day. A conventional cryoplant, consisting of a cold box and a set of primary heat exchangers, is ill-suited to such a mode of operation as the instantaneou...

  3. 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.

  4. Measurements of plasma position in TJ-I Tokamak

    International Nuclear Information System (INIS)

    Qin, J.; Ascasibar, E.; Navarro, A.P.; Ochando, M.A.; Pastor, I.; Pedrosa, M.A.; Rodriguez, L.; Sanchez, J.; Team, TJ-I.

    1994-01-01

    This report presents the experimental measurements of plasma position in TJ-I tokamak by using small magnetic probes. The basis of method has been described in our previous work (1) in which the plasma current is considered as a filament current. The observed relations between the disruptive instabilities and plasma displacements are also show here. (Author) 7 refs

  5. Extraordinary mode absorption at the electron cyclotron harmonic frequencies as a Tokamak plasma diagnostic

    International Nuclear Information System (INIS)

    Pachtman, A.

    1986-09-01

    Measurements of Extraordinary mode absorption at the electron cyclotron harmonic frequencies are of unique value in high temperature, high density Tokamak plasma diagnostic applications. An experimental study of Extraordinary mode absorption at the semi-opaque second and third harmonics has been performed on the ALCATOR C Tokamak. A narrow beam of submillimeter laser radiation was used to illuminate the plasma in a horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission

  6. Langmuir probe evaluation of the plasma potential in tokamak edge plasma for non-Maxwellian EEDF

    Energy Technology Data Exchange (ETDEWEB)

    Popov, Ts.K. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Dimitrova, M. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Ivanova, P. [Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Hasan, E. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Horacek, J.; Dejarnac, R.; Stoeckel, J.; Weinzettl, V. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Kovacic, J. [Jozef Stefan Institute, Ljubljana (Slovenia)

    2014-04-15

    The First derivative probe technique for a correct evaluation of the plasma potential in the case of non-Maxwellian EEDF is presented and used to process experimental data from COMPASS tokamak. Results obtained from classical and first derivative techniques are compared and discussed. The first derivative probe technique provides values for the plasma potential in the scrape-off layer of tokamak plasmas with an accuracy of about ±10%. Classical probe technique can provide values of the plasma potential only, if the electron and ion temperatures are known as well as the coefficient of secondary electron emission. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  7. Multi-channel bolometer system on JFT-2M tokamak

    International Nuclear Information System (INIS)

    Tamai, Hiroshi; Maeno, Masaki; Matsuda, Toshiaki; Matoba, Tohru

    1988-07-01

    Multi-channel bolometer system is designed and installed to observe the radiation profile on JFT-2M tokamak. Sensor head is made of Thinistor, which is a kind of semiconductor, because it has the advantage of higher sensitivity of about one order of magnitude than the conventional metal foil bolometer and is suitable for the profile measurement in which the signal from the plasma is relatively small. The response and cooling characteristics of the bolometer sensor are suitable for the condition of JFT-2M tokamak plasma. Low noise circuit of bridge and differentiator is developed to optimize the signal to noise ratio in the JFT-2M operating condition. With use of the bolometer system, the radiation profile in joule heating plasma as well as additional heating plasma especially in H-mode plasma is successfully observed. (author)

  8. Can tokamaks PFC survive a single event of any plasma instabilities?

    Science.gov (United States)

    Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

    2013-07-01

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

  9. Can tokamaks PFC survive a single event of any plasma instabilities?

    International Nuclear Information System (INIS)

    Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

    2013-01-01

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time

  10. Can tokamaks PFC survive a single event of any plasma instabilities?

    Energy Technology Data Exchange (ETDEWEB)

    Hassanein, A., E-mail: hassanein@purdue.edu [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States); Sizyuk, V.; Miloshevsky, G.; Sizyuk, T. [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

    2013-07-15

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

  11. One dimensional simulation on stability of detached plasma in a tokamak divertor

    International Nuclear Information System (INIS)

    Nakazawa, Shinji; Nakajima, Noriyoshi; Okamoto, Masao; Ohyabu, Nobuyoshi

    1999-06-01

    The stability of radiation front in the Scrape-Off-Layer (SOL) of a tokamak is studied with a one dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. The simulation results show that stable detached solutions exist, where the plasma temperature near the divertor target is ∼2 eV. It is found that whenever such stable detached states are attained, the strong radiation front is contact with or at a small distance from the divertor target. When the energy externally injected into the SOL is decreased below a critical value, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such cases, no stationary solutions such that the radiation front rests in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of radiation front. (author)

  12. Hybrid model for simulation of plasma jet injection in tokamak

    Science.gov (United States)

    Galkin, Sergei A.; Bogatu, I. N.

    2016-10-01

    Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

  13. The role of high speed photography in plasma instability research on the AEC tokamak

    International Nuclear Information System (INIS)

    Fletcher, J.D.; Coster, D.P.; De Villiers, J.A.M.; Kotze, P.B.; Nothnagel, G.; O'Mahony, J.R.; Roberts, D.E.; Sherwell, D.

    1986-01-01

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions in fusion research devices like tokamaks. Such a system has been installed on the AEC tokamak. This paper reports some preliminary results obtained during typical plasma discharges

  14. Research on the penetration characteristic of supersonic molecule beam on the HL-1M tokamak

    International Nuclear Information System (INIS)

    Dong Jiafu; Luo Junling; Tang Nianyi; Li Wei; Zhong Yunze; Liu Yi; Fu Binzhong; Yao Lianghua; Feng Baibing; Qin Yunwen

    2002-01-01

    In the HL-1M plasma experiments, two refueling ways, the pellet injection and the gas puffing, are usually used. In recent years, a new refueling method, the supersonic molecular beam (SMB) injection, has been developed. In this paper, the penetration characteristics of SMB has been analyzed and researched. The SMB experiments were analyzed with H α emissivity measurement. The H α emissivity distribution in the plasma was calculated from multi-channel chord emission singles by means of the asymmetry Abel inversion method

  15. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  16. Electron density and temperature determination in a Tokamak plasma using light scattering

    International Nuclear Information System (INIS)

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

    1976-01-01

    A theoretical foundation review for light scattering by plasmas is presented. Furthemore, a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, is included in a Tokamak plasma using spectral analysis of the scattered radiation. (author) [es

  17. 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.

  18. Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak

    International Nuclear Information System (INIS)

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

    1994-01-01

    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

  19. A model for plasma discharges simulation in Tokamak devices

    International Nuclear Information System (INIS)

    Fonseca, Antonio M.M.; Silva, Ruy P. da; Galvao, Ricardo M.O.; Kusnetzov, Yuri; Nascimento, I.C.; Cuevas, Nelson

    2001-01-01

    In this work, a 'zero-dimensional' model for simulation of discharges in Tokamak machine is presented. The model allows the calculation of the time profiles of important parameters of the discharge. The model was applied to the TCABR Tokamak to study the influence of parameters and physical processes during the discharges. Basically it is constituted of five differential equations: two related to the primary and secondary circuits of the ohmic heating transformer and the other three conservation equations of energy, charge and neutral particles. From the physical model, a computer program has been built with the objective of obtaining the time profiles of plasma current, the current in the primary of the ohmic heating transformer, the electronic temperature, the electronic density and the neutral particle density. It was also possible, with the model, to simulate the effects of gas puffing during the shot. The results of the simulation were compared with the experimental results obtained in the TCABR Tokamak, using hydrogen gas

  20. 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)

  1. Plasma diagnostics using synchrotron radiation in tokamaks

    International Nuclear Information System (INIS)

    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

  2. Control of tokamak plasma current and equilibrium with hybrid poloidal field coils

    International Nuclear Information System (INIS)

    Shimada, Ryuichi

    1982-01-01

    A control method with hybrid poloidal field system is considered, which comprehensively implements the control of plasma equilibrium and plasma current, those have been treated independently in Tokamak divices. Tokamak equilibrium requires the condition that the magnetic flux function value on plasma surface must be constant. From this, the current to be supplied to each coil is determined. Therefore, each coil current is the resultant of the component related to plasma current excitation and the component required for holding equilibrium. Here, it is intended to show a method by which the current to be supplied to each coil can easily be calculated by the introduction of hybrid control matrix. The text first considers the equilibrium of axi-symmetrical plasma and the equilibrium magnetic field outside plasma, next describes the determination of current using the above hybrid control matrix, and indicates an example of controlling Tokamak plasma current and equilibrium by the hybrid poloidal field coils. It also shows that the excitation of plasma current and the maintenance of plasma equilibrium can basically be available with a single power supply by the appropriate selection of the number of turns of each coil. These considerations determine the basic system configuration as well as decrease the installed capacity of power source for the poloidal field of a Tokamak fusion reactor. Finally, the actual configuration of the power source for hybrid poloidal field coils is shown for the above system. (Wakatsuki, Y.)

  3. Plasma vertical instability in a tokamak with rail limiters

    International Nuclear Information System (INIS)

    Belashob, V.I.; Brevnov, N.N.; Gribov, Yu.V.; Putvinskij, S.V.

    1989-01-01

    An effect of currents between rail limiters on plasma equilibrium in the tokamak is studied theoretically and experimentally. Limiter currents can emerge at fast changes of plasma position along rail limiters for example when compression along major radius takes place and result in additional electrodynamic loadings on to the chamber and limiters. It is shown that at high currents between the limiters, the behaviour of discharge depends on limiter voltage polarity. When the plasma - limiter contact points are asymmetrically located respective to an equatorial plane a radial component of the limiter current emerges. The interaction of the component with the toroidal magnetic field can result in a vertical plasma filament instability. 9 refs.; 10 figs

  4. Nanoparticle Plasma Jet as Fast Probe for Runaway Electrons in Tokamak Disruptions

    Science.gov (United States)

    Bogatu, I. N.; Galkin, S. A.

    2017-10-01

    Successful probing of runaway electrons (REs) requires fast (1 - 2 ms) high-speed injection of enough mass able to penetrate through tokamak toroidal B-field (2 - 5 T) over 1 - 2 m distance with large assimilation fraction in core plasma. A nanoparticle plasma jet (NPPJ) from a plasma gun is a unique combination of millisecond trigger-to-delivery response and mass-velocity of 100 mg at several km/s for deep direct injection into current channel of rapidly ( 1 ms) cooling post-TQ core plasma. After C60 NPPJ test bed demonstration we started to work on ITER-compatible boron nitride (BN) NPPJ. Once injected into plasma, BN NP undergoes ablative sublimation, thermally decomposes into B and N, and releases abundant B and N high-charge ions along plasma-traversing path and into the core. We present basic characteristics of our BN NPPJ concept and first results from B and N ions on Zeff > 1 effect on REs dynamics by using a self-consistent model for RE current density. Simulation results of BNQ+ NPPJ penetration through tokamak B-field to RE beam location performed with Hybrid Electro-Magnetic code (HEM-2D) are also presented. Work supported by U.S. DOE SBIR Grant.

  5. Electron density and temperature determination in a Tokamak plasma using light scattering

    International Nuclear Information System (INIS)

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

    1976-01-01

    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

  6. Total hydrogen and oxygen fluxes in the edge plasma of tokamaks

    International Nuclear Information System (INIS)

    Kastelewicz, H.

    1988-01-01

    A relativistic model of the edge plasma of tokamaks is described considering the primary neutral fluxes emitted from limiter and wall. The primary neutrals, which determine essentially the particle flux balance in the plasma edge, the scrape-off layer plasma and the particles adsorbed at limiter and wall are treated as separate subsystems which are iteratively coupled through the mutual particle sinks and sources. The model is used for the calculation of total hydrogen and oxygen fluxes in edge plasma of tokamaks. The results for different fractions of and contributions to the total fluxes are illustrated and discussed

  7. Properties of the tokamak edge plasma

    International Nuclear Information System (INIS)

    Wolff, H.

    1988-01-01

    A short review of some features of the edge plasma in limiter tokamaks is given. The limits of the simple one-dimensional scrape-off layer (SOL) model and the relation between the core plasma are discussed. Multifaceted asymmetric radiation from the edge (MARFE) phenomena and detached plasma are closely connected with the particle and energy balance of the SOL. Their occurrence is based on the relation of plasma parameters of the edge plasma to those of the core. Important problems of plasma wall interactions are the detection of the impurity sources and sinks and the study of the impurity transport and shielding. The non-uniform character of plasma wall interactions and their dependence on the discharge performance still renders difficult any theoretical forecast of impurity distribution and transport and calls for better diagnostics. (author)

  8. Plasma residual rotation in the TCABR tokamak

    International Nuclear Information System (INIS)

    Severo, J.H.F.; Nascimento, I.C.; Tsypin, V.S.; Galvao, R.M.O.

    2003-01-01

    This paper reports the first results on the measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak, in the collisional regime (Pfirsch-Schluter), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5±1.0) x 10 3 m s -1 at r ∼ 2/3a,(a-limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) x 10 3 m s -1 , at the plasma core, to a co-current value of (2.0 ± 0.9) x 10 3 m s -1 near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner (1991 Phys. Fluids B 3 2050). (author)

  9. Probing of the peripheral plasma of a tokamak by diaphragm thermography

    International Nuclear Information System (INIS)

    Gyunter, K.; Grunov, K.; Evstigneev, S.A.

    1985-01-01

    The tokamak limiter diverts significant particle and energy fluxes from the plasma and sharply decreases the interaction between the plasma and wall, thus protecting the wall from destruction and the plasma from catastrophic contamination. The temperature of a moveble limiter is measured in the device on the basis of the infrared radiation from its surface. By applying certain models of the layer near the wall this permits to calculate the magnitude of the heat flux directed to the limiter. The power density is, in normal discharges, 2-3 kW/cm 2 ; in discharges with disruptions the power density may reach 10 kW/cm 2 . The measurements show that longitudinal energy flux in the peripheral layer of the plasma filament decreases radially in an exponential manner and can yield the parameter of the decrease. It is found that during discharges involving breakdown a very nonuniform heating of the limiter surface occurs at the initial and final stages of normal discharges, this indicates the existence of a certain structure of the peripheral plasma layer

  10. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

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

    International Nuclear Information System (INIS)

    Casati, A.

    2009-10-01

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

  12. Magnetic spires for the detection of the position of the plasma column in a Tokamak (linear approximation)

    International Nuclear Information System (INIS)

    Colunga S, S.

    1990-07-01

    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)

  13. 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/{{\

  14. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

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

  15. Tokamak-FED plasma-engineering assessments

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Lyon, J.F.; Rutherford, P.H.

    1981-01-01

    A wide range of plasma assumptions and scenarios has been examined for the current US tokamak FED concept, which aims to provide a controlled, long pulse (approx. 100 s) burning plasma with an energy amplification of greater than or equal to 5, a fusion power of 180 MW, and a neutron wall load of greater than or equal to 0.4 MW/m 2 . The results of the assessment suggest that the current FED baseline parameters of R = 5.0 m, B/sub t/ = 3.6 T, a = 1.3 m, b = 2.1 m (D-shape), and I/sub p/ = 5.4 MA are appropriate in reaching the above plasma performance, despite uncertainties in several plasma physics areas, such as confinement scaling, achievable beta, impurity control, etc. To enhance the probability of achieving fusion ignition and to provide some margin against a short fall in our physics projections in FED, a limited operating capability at B/sub t/ = 4.6 T and I/sub p/ = 6.5 MA is incorporated. Various other options and remedies have also been assessed aiming to alleviate the impact of the uncertainties on the FED design concept. These approaches appear promising because they can be studied within the current fusion physics program and may lead to drastically more cost-effective FED concepts

  16. Planned upgrade to the coaxial plasma source facility for high heat flux plasma flows relevant to tokamak disruption simulations

    International Nuclear Information System (INIS)

    Caress, R.W.; Mayo, R.M.; Carter, T.A.

    1995-01-01

    Plasma disruptions in tokamaks remain serious obstacles to the demonstration of economical fusion power. In disruption simulation experiments, some important effects have not been taken into account. Present disruption simulation experimental data do not include effects of the high magnetic fields expected near the PFCs in a tokamak major disruption. In addition, temporal and spatial scales are much too short in present simulation devices to be of direct relevance to tokamak disruptions. To address some of these inadequacies, an experimental program is planned at North Carolina State University employing an upgrade to the Coaxial Plasma Source (CPS-1) magnetized coaxial plasma gun facility. The advantages of the CPS-1 plasma source over present disruption simulation devices include the ability to irradiate large material samples at extremely high areal energy densities, and the ability to perform these material studies in the presence of a high magnetic field. Other tokamak disruption relevant features of CPS-1U include a high ion temperature, high electron temperature, and long pulse length

  17. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  18. Study of plasma discharge evolution and edge turbulence with fast visible imaging in the Aditya tokamak

    International Nuclear Information System (INIS)

    Banerjee, Santanu; Manchanda, R.; Chowdhuri, M.B.

    2015-01-01

    Study of discharge evolution through the different phases of a tokamak plasma shot viz., the discharge initiation, current ramp-up, current flat-top and discharge termination, is essential to address many inherent issues of the operation of a Tokamak. Fast visible imaging of the tokamak plasma can provide valuable insight in this regard. Further, edge turbulence is considered to be one of the quintessential areas of tokamak research as the edge plasma is at the immediate vicinity of the plasma core and plays vital role in the core plasma confinement. The edge plasma also bridges the core and the scrape off layer (SOL) of the tokamak and hence has a bearing on the particle and heat flux escaping the plasma column. Two fast visible imaging systems are installed on the Aditya tokamak. One of the system is for imaging the plasma evolution with a wide angle lens covering a major portion of the vacuum vessel. The imaging fiber bundle along with the objective lens is installed inside a radial re-entrant viewport, specially designed for the purpose. Another system is intended for tangential imaging of the plasma column. Formation of the plasma column and its evolution are studied with the fast visible imaging in Aditya. Features of the ECRH and LHCD operations on Aditya will be discussed. 3D filaments can, be seen at the plasma edge all along the discharge and they get amplified in intensity at the plasma termination phase. Statistical analysis of these filaments, which are essentially plasma blobs will be presented. (author)

  19. Automation of Aditya tokamak plasma position control DC power supply

    Energy Technology Data Exchange (ETDEWEB)

    Arambhadiya, Bharat, E-mail: bharat@ipr.res.in; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.

    2016-11-15

    Highlights: • Plasma position control is very essential for obtaining repeatable high temperature, high-density discharges of longer durations in tokomak. • The present capacitor bank has limitations of maximum current capacity and position control beyond 200 ms. • The installation of a separate set of coils and a DC power supply can control the plasma position beyond 200 ms. • A high power thyristor (T588N1200) triggers for DC current pulse of 300 A fires precisely at required positions to modify plasma position. • The commissioning is done for the automated in-house, quick and reliable solution. - Abstract: Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200 ms for position control. This motivated to have a DC power supply of 500 A to provide current pulse beyond 200 ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.

  20. Von Hippel-Lindau disease (vHL)

    DEFF Research Database (Denmark)

    Binderup, Marie Louise Mølgaard; Bisgaard, Søs Marie Luise; Harbud, Vibeke

    2013-01-01

    for vHL which is comprised of Danish doctors and specialists interested in vHL. The recommendations are based on longstanding clinical experience, Danish original research, and extensive review of the international literature. vHL is a hereditary multi-tumour disease caused by germline mutations....../MRI of the abdomen, e) annual plasma-metanephrine, plasma-normetanephrine, and plasma-chromogranin A tests, and f) annual hearing examination at a department of audiology. It is advised that one doctor takes on the responsibility of coordination of and referral to the many examinations, and the communication...

  1. A study on the heating and diagnostic of a tokamak plasma by electromagnetic waves of the electron cyclotron range of frequencies

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi

    1989-09-01

    A study on the heating and diagnosis of tokamak plasma by electromagnetic waves of electron cyclotron range of frequency is summarized. The main results obtained are as follows. On the engineering and technology, the technology of injecting high frequency, large power millimeter waves into tokamak plasma was established by carrying out the design, manufacture and test of a 60 GHz, 400 kW high frequency heating system, and the design, manufacture and test of a heterodyne type electron cyclotron radiation multi-channel mealsuring system were carried out, and the technology of measuring the radiation from tokamak plasma with the time resolution of 10 μs in multi-channel was established. On nuclear fusion reactor core engineering and plasma physics, the high efficiency electron heating of tokamak plasma by the incidence of fundamental irregular and regular waves at electron cyclotron frequency was verified. The discovery and analysis of the heating by electrostatic waves arising due to mode transformation from electromagnetic waves in upper hybrid resonance layer were carried out. By the incidence of second harmonic waves, the high efficiency electron heating of tokamak plasma was verified, and the heating characteristics were clarified. And others. (K.I.) 179 refs

  2. Negative edge plasma currents in the SINP tokamak

    Indian Academy of Sciences (India)

    RAE is the maximum runaway energy emitted during a burst period of tdur. HXR. There being no plasma control feedback system in the SINP tokamak, the dynamics of the plasma equilibrium is time-dependent and the column shift is now made by the discharge dynamics itself. We measured DRAE for the two discharges ...

  3. TFTR/JET INTOR workshop on plasma transport tokamaks

    International Nuclear Information System (INIS)

    Singer, C.E.

    1985-01-01

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

  4. Eight-shot pellet injector and fueling experiments at the HL-1M tokamak

    International Nuclear Information System (INIS)

    Xiao Zhenggui; Li Bo; Li Li

    2001-01-01

    An Eight-shot Pellet Injection (EPI) system has been proposed and developed in collaboration between STU (St. Petersburg State Technical University) of Russia and SWIP. In the EPI, the I n-situ c ondensation technique was used to produce the pellets in eight gun barrels respectively. The nominal pellet size (diameter of 1.0 mm and of 1.4 mm or 1.2 mm) is limited by the gun barrel inner diameter. The pellet length is adjusted by changing the g radient temperature o n the gun barrels and the amounts of filling fuel gas. Pellets are fired at speed range of 200 - 1200 m/s by He propellant with pressure of 2 - 6 MPa and then transferred to HL-1M vessel through an injection line that consists of two set of differential vacuum pumped chambers and guide tube combined with fast valves. In addition, this unit is equipped with diagnostics for pellet velocity and shape measure. The EPI has installed on HL-1M since 1996 for the multi-shot pellet fueling experiments. The typical characteristics including the peaked density profile and improved confinement, the deep penetration and suppression of soft X-ray sawteeth, the variance of rotation and flow of plasma in edge region as well as the photographing of pellet ablation clouds are presented

  5. ICRF heating experiments in JFT-2 tokamak

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1986-01-01

    This is an experimental study of ICRF heating on JFT-2 Tokamak in Japan Atomic Energy Research Institute. In this study, we first clarified physical and engineering problems of ICRF heating of tokamak plasma. Next, we optimized the design of the ICRF heating system, and the plasma parameters for the heating. Finally, we could demonstrate a high efficiency of this additional heating method by launching RF power which is two or three times as large as an ohmic input power to a plasma. And we achieved following things. (1) We optimized a design of an antenna, and we improved a durability of the system for high voltage. With the result that we achieved the maximum power density on an antenna. (2) We demonstrated that electron heating regime and ion heating regime can be easily accessed by controlling plasma parameters. Also we found the optimum heating conditions in each heating regime. (3) We experimentally clarified the production mechanism of impurities during ICRF heating. We could reduce the influx of metal impurity ions to a plasma by employing low z materials for limiters and antenna shields. Consequently, we improved a heating efficiency of electrons. Next, we studied a power balance of plasma during ICRF heating, and we could compare heating characteristics of ICRF with other additional heatings on JFT-2. (author)

  6. Internal helical modes with m > 1 in a tokamak with a small shear and high plasma pressure

    International Nuclear Information System (INIS)

    Mikha lovskij, A.B.; Aburdzhaniya, G.D.; Krymskij, A.M.

    1979-01-01

    Internal helical modes with m>1 in a circular cross-section tokamak with a small shear and large value of the parameter β (β is the ratio between the mean plasma pressure and the mean pressure of the poloidal magnetic field) are investigated. The equations obtained are used to study the destabilizing effects leading to helical instabilities. The role of destabilizing effects is regarded both in local and in a nonlocal approximations on the assumption that the radial plasma pressure is distributed parabolically and that the radial current distribution is also parabolic though slightly varying. It has been established that the profiling of current may lead to the tokamak plasma stability with respect to the modes under investigation. A tokamak with a small shear has been shown to be more stable relative to these modes than that with a large shear

  7. A study on tokamak fusion reactor - Numerical analyses of MHD equilibrium= and edge plasma transport in tokamak fusion reactor with divertor configurations

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Lim, Ki Hang; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Duk Kyu [Seoul National University, Seoul (Korea, Republic of); Cho, Soo Won [Kyungki Unviersity, Suwon (Korea, Republic of)

    1995-08-01

    In the present project for developing the numerical codes of 2-DMHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we compute the plasma equilibrium of double null type and calculate the external coil currents and the plasma parameters used for operation and control data. Also the numerical algorithm is developed to analyse the behavior of edge plasmas in poloidal and radial directions and the programming and debugging of a 2-D transport code are completed. Furthermore, a neutral particle transport code for the edge region is developed and then used for the analysis of the neutral transport phenomena giving the sources in the fluid equations, and expected to supply the input parameters for the edge plasma transport code. 34 refs., 5 tabs., 28 figs. (author)

  8. Plasma density determination by microwave interferometry. The 2 mm interferometer of the TJ-1 tokamak

    International Nuclear Information System (INIS)

    Manero, F.; Martin, R.

    1984-01-01

    In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density profiles from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (author)

  9. Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak

    International Nuclear Information System (INIS)

    Martin, R.; Manero, F.

    1984-01-01

    In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs

  10. Scrape-off layer plasma modeling for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Porter, G.D.; Rognlien, T.D.; Allen, S.L.

    1994-09-01

    The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing

  11. Plasma current sustainment after iron core saturation in the STOR-M tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Mitarai, O., E-mail: omitarai@ktmail.tokai-u.jp [Kumamoto Liberal Arts Education Center, Tokai University, 9-1-1 Toroku, Higashi-ku, Kumamoto 862-8652 (Japan); Ding, Y.; Hubeny, M.; Lu, Y.; Onchi, T.; McColl, D.; Xiao, C.; Hirose, A. [Plasma Physics Laboratory, University of Saskatchewan, 116 Science Place, Saskatoon, SK S7N 5E2 (Canada)

    2014-10-15

    Highlights: • Plasma current can be started up by small iron core without central solenoid. • Iron core removes central solenoid. • Plasma current can be maintained after iron core saturation. • Hysteresis curve shows the partial core saturation. • Image field from iron core is estimated during discharge. • Spherical tokamak reactor without CS is proposed using the small iron core. - Abstract: We propose to use of a small iron core transformer to start up the plasma current in a spherical tokamak (ST) reactor without central solenoid (CS). Taking advantage of the high aspect ratio of the STOR-M iron core tokamak, we have demonstrated that the plasma current up to 10–15 kA can be started up using the outer Ohmic heating (OH) coils without CS, and that the plasma current can be maintained further by increasing the outer OH coil current during iron core saturation phase. When the magnetizing current reaches 1.2 kA and the iron core becomes saturated, the third capacitor bank connected to the outer OH coils is discharged to maintain the plasma current. The plasma current is slightly increased and maintained for additional 5 ms as expected from numerical calculations. Core saturation has been clearly observed on the hysteresis curve. This is the first experimental demonstration of the feasibility of slow transition from the iron core to air core transformer phase without CS. The results implies that a plasma current can be initiated by a small iron core and could be ramped up by additional heating and vertical field after iron core saturation in future STs without CS.

  12. A Toroidally Symmetric Plasma Simulation code for design of position and shape control on tokamak plasmas

    International Nuclear Information System (INIS)

    Takase, Haruhiko; Senda, Ikuo

    1999-01-01

    A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)

  13. I.R. and F.I.R. laser polarimetry as a diagnostic tool in high-β and Tokamak plasmas

    International Nuclear Information System (INIS)

    Pereira, D.; Machida, M.; Scalabrin, A.

    1986-01-01

    The change of the polarization state of an electromagnetic wave (E.M.W.) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the E.M.W. allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-Pinches plasmas (lambda approx. 10μm). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-Pinch/UNICAMP, using lasers developed at UNICAMP. (Author) [pt

  14. Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak; Determinacion de la densidad de un plasma por interferometria de microondas. El interferometro de 2 mm del Tokamak TJ-1

    Energy Technology Data Exchange (ETDEWEB)

    Martin, R; Manero, F

    1984-07-01

    In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs.

  15. Stabilization of external kink modes in a tokamak with rotating plasma

    International Nuclear Information System (INIS)

    Mikhailovskii, A.B.; Kuvshinov, B.N.

    1995-01-01

    An analytical theory of stabilization of external kink modes in a tokamak with rotating plasma is developed, which is of interest in connection with experiments on the DIII-D tokamak demonstrating such a stabilization. It is assumed that, in addition to the main poloidal harmonic, the mode includes one or more side-band poloidal harmonics with singular points lying inside the plasma. Near these singular points, plasma inertia and related toroidal effects, the compressible part of plasma pressure and longitudinal viscosity, are allowed for. These effects are described kinetically taking into account the toroidal trapping of the resonant ions, which is essential if the toroidal velocity is small compared to the ion thermal velocity. Thereby, the theory presented includes both ion Landau damping and its weakening due to toroidal trapping. Near the singular points high-beta effects, which result in the finiteness of the Mercier index s, are allowed for. It is shown that the influence of plasma rotation on the external kink modes is most significant in the case of s<0, i.e., when the development of the instability in a non-rotating plasma is most highly favored. In this case, the plasma rotation plays a stabilizing role, even when the ion Landau damping is neglected. The analysis presented also confirms the hypothesis of Bondeson and Ward on the stabilizing effect of ion Landau damping if this damping is not too small

  16. An investigation on detection and measurement of fusion neutron spectrum and radiation flux in large tokamak

    International Nuclear Information System (INIS)

    Yang Jinwei; Li Wenzhong; Zhang Wei

    2003-01-01

    The detection methods, detectors and spectrometers of D-D and D-T fusion neutron have been overviewed in large tokamaks. Some options are proposed for developing new detection systems of fusion neutrons suitable to the HL-2A tokamak. (authors)

  17. Robust control design for the plasma horizontal position control on J-TEXT Tokamak

    International Nuclear Information System (INIS)

    Yu, W.Z.; Chen, Z.P.; Zhuang, G.; Wang, Z.J.

    2013-01-01

    It is extremely important for tokamak to control the plasma position during routine discharge. However, the model of plasma in tokamak usually contains much of the uncertainty, such as structured uncertainties and unmodeled dynamics. Compared with the traditional PID control approach, robust control theory is more suitable to handle this problem. In the paper, we propose a H ∞ robust control scheme to control the horizontal position of plasma during the flat-top phase of discharge on Joint Texas Experimental Tokamak (J-TEXT) tokamak. First, the model of our plant for plasma horizontal position control is obtained from the position equilibrium equations. Then the H ∞ robust control framework is used to synthesize the controller. Based on this, an H ∞ controller is designed to minimize the regulation/tracking error. Finally, a comparison study is conducted between the optimized H ∞ robust controller and the traditional PID controller in simulations. The simulation results of the H ∞ robust controller show a significant improvement of the performance with respect to those obtained with traditional PID controller, which is currently used on our machine

  18. Neutral-beam requirements for compression-boosted ignited tokamak plasmas

    International Nuclear Information System (INIS)

    Cohn, D.R.; Jassby, D.L.; Kreischer, K.

    1977-12-01

    Neutral-beam energies of 200 to 500-keV D 0 may be required to insure adequate penetration into the center of ignition-sized tokamak plasmas. However, the beam energy requirement can be reduced by using a start-up scenario in which the final plasma is formed by major-radius compression of a beam-heated plasma whose density-radius product, na, is determined by satisfactory neutral-beam penetration. ''Compression boosting'' is attractive only for plasmas in which ntau/sub E/ increases with na, because a major-radius compression C increases na by C 3 / 2 . The dependence on C of beam energy and beam power for plasmas which obey ''empirical scaling laws'' of the type ntau/sub E/ varies as (na) 2 is analyzed. The dependences on C of stored magnetic energy and TF-coil power dissipation are also determined. It is found that a compression ratio of 1.5 to attain the ignited plasma permits adequate penetration by 150-keV D 0 beams

  19. DIII-D Integrated plasma control solutions for ITER and next-generation tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Ferron, J.R.; Hyatt, A.W.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; In, Y.

    2008-01-01

    Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed 'integrated plasma control,' enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple ('control-level') models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks

  20. Development of plasma diagnostics technologies - Measurement of transport= parameters in tokamak edge plasma by using electric transport probes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kyu Sun; Chang, Do Hee; Sim, Yeon Gun; Kim, Jin Hee [Hanyang University, Seoul (Korea, Republic of)

    1995-08-01

    Electric transport probe system is developed for the measurement of electron temperature, floating potential, plasma density and flow velocity of= edge plasmas in the KT-2 medium size tokamak. Experiments have been performed in KT-1 small size tokamak. Electric transport probe is composed of a single probe(SP) and a Mach probe (MP). SP is used for the measurements of electron density, floating potential, and plasma density and measured values are {approx} 3*10{sup 11}/cm{sup -3}, -20 volts, 15 {approx} 25 eV. For the most discharges, respectively. MP is for the measurements of toroidal(M{sub T}) and poloidal(M{sub P}) flow velocities, and density, which are M{sub T} {approx_equal} .0.85, M{sub P} {approx_equal}. 0.17, n. {approx_equal} 2.1*10{sup 11} cm{sup -3}, respectively. A triple probe is also developed for the direct reading of T{sub e} and n{sub e}, and is used for DC, RF, and RF+DC plasma in APL of Hanyang university. 38 refs., 36 figs. (author)

  1. The major tokamak distruption in cylindrical plasma

    International Nuclear Information System (INIS)

    Choi, Jeong Sik; Choi, Eun Ha; Choi, Duk In

    1986-01-01

    The mechanism of the major disruption in tokamak plasma which involves the nonlinear interaction of tearing models is numerically studied in two and three dimensional formulations. In this study, it is found that in the two dimensional case with a flattened current density profile the magnetic islands of the m=2; n=1 mode do not saturate nonlinearly and but strongly interact with the limiter. Thus it is suggested that the helical perturbation of the m=2;n=1 mode plays the dominant role in the major disruption. We also show that the m=2;n=1 mode nonlinearly destablizes other tearing modes, especially the m=3;n=2 mode, from the nonlinear coupling of different helicities as also shown in other studies. The plasma extends across the plasma cross section, and the plasma core shifts inward along the major radius during the major disruption. The numerical result for the major disruption time measured using the nonlinear 3-D procedure for the initial value problem with PLT parameters is about 450 μsec which agrees reasonably well with the experimental value of 500 μsec. (Author)

  2. Electron and current density measurements on tokamak plasmas

    International Nuclear Information System (INIS)

    Lammeren, A.C.A.P. van.

    1991-01-01

    The first part of this thesis describes the Thomson-scattering diagnostic as it was present at the TORTUR tokamak. For the first time with this diagnostic a complete tangential scattering spectrum was recorded during one single laser pulse. From this scattering spectrum the local current density was derived. Small deviations from the expected gaussian scattering spectrum were observed indicating the non-Maxwellian character of the electron-velocity distribution. The second part of this thesis describes the multi-channel interferometer/ polarimeter diagnostic which was constructed, build and operated on the Rijnhuizen Tokamak Project (RTP) tokamak. The diagnostic was operated routinely, yielding the development of the density profiles for every discharge. When ECRH (Electron Cyclotron Resonance Heating) is switched on the density profile broadens, the central density decreases and the total density increases, the opposite takes place when ECRH is switched off. The influence of MHD (magnetohydrodynamics) activity on the density was clearly observable. In the central region of the plasma it was measured that in hydrogen discharges the so-called sawtooth collapse is preceded by an m=1 instability which grows rapidly. An increase in radius of this m=1 mode of 1.5 cm just before the crash is observed. In hydrogen discharges the sawtooth induced density pulse shows an asymmetry for the high- and low-field side propagation. This asymmetry disappeared for helium discharges. From the location of the maximum density variations during an m=2 mode the position of the q=2 surface is derived. The density profiles are measured during the energy quench phase of a plasma disruption. A fast flattening and broadening of the density profile is observed. (author). 95 refs.; 66 figs.; 7 tabs

  3. Real-time control of Tokamak plasmas: from control of physics to physics-based control

    International Nuclear Information System (INIS)

    Felici, F. A. A.

    2011-11-01

    Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a digital realtime control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This experimental platform has been used to develop and test new control strategies for three plasma physics instabilities: sawtooth, edge localized mode (ELM) and neoclassical tearing mode (NTM). The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface (q: safety factor). A sawtooth pacing controller was developed which is able to control the time of appearance of the next sawtooth crash. Each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes have been used to study the relationship between sawteeth and NTMs. Postcrash MHD activity can provide the ‘seed’ island for an NTM, which then grows under its neoclassical bootstrap drive. The seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have

  4. Determination of plasma column transverse section in the TBR-1 tokamak

    International Nuclear Information System (INIS)

    Conde, M.E.

    1986-01-01

    The temporal evolution of plasma column transverse section in the TBR-1 tokamak is determined. The experimental melhod is based on the simulation of toroidal current distribution in plasma by a set of toroidal filaments. The currents in these filaments are determined by minimization of square error between the magnetic field produced by filaments and the field measured into the tokamak vacuum vessel. For the measurement of magnetic field, twenty small magnetic coils were constructed and installated in the region protected by current limiters. The plasma column transverse cross section is determined by poloidal field produced by the currents in filaments. The multipole moments of plasma current distribution and the Λ Shafranov parameter were obtained. (M.C.K.) [pt

  5. A general comparison between tokamak and stellarator plasmas

    Directory of Open Access Journals (Sweden)

    Yuhong Xu

    2016-07-01

    Full Text Available This paper generally compares the essential features between tokamaks and stellarators, based on previous review work individually made by authors on several specific topics, such as theories, bulk plasma transport and edge divertor physics, along with some recent results. It aims at summarizing the main results and conclusions with regard to the advantages and disadvantages in these two types of magnetic fusion devices. The comparison includes basic magnetic configurations, magnetohydrodynamic (MHD instabilities, operational limits and disruptions, neoclassical and turbulent transport, confinement scaling and isotopic effects, plasma rotation, and edge and divertor physics. Finally, a concept of quasi-symmetric stellarators is briefly referred along with a comparison of future application for fusion reactors.

  6. Investigation of small-scale tokamak plasma turbulence by correlative UHR backscattering diagnostics

    International Nuclear Information System (INIS)

    Gusakov, E Z; Gurchenko, A D; Altukhov, A B; Bulanin, V V; Esipov, L A; Kantor, M Yu; Kouprienko, D V; Lashkul, S I; Petrov, A V; Stepanov, A Yu

    2006-01-01

    Fine scale turbulence is considered nowadays as a possible candidate for the explanation of anomalous ion and electron energy transport in magnetized fusion plasmas. The unique correlative upper hybrid resonance backscattering (UHR BS) technique is applied at the FT-2 tokamak for investigation of density fluctuations excited in this turbulence. The measurements are carried out in Ohmic discharge at several values of plasma current and density and during current ramp up experiment. The moveable focusing antennas set have been used in experiments allowing probing out of equatorial plane. The radial wave number spectra of the small-scale component of tokamak turbulence are determined from the correlation data with high spatial resolution. Two small-scale modes possessing substantially different phase velocities are observed in plasma under conditions when the threshold for the electron temperature gradient mode excitation is overcome. The possibility of plasma poloidal velocity profile determination using the UHR BS signal is demonstrated

  7. Trade studies of plasma elongation for next-step tokamaks

    International Nuclear Information System (INIS)

    Galambos, J.D.; Strickler, D.J.; Peng, Y.K.M.; Reid, R.L.

    1988-09-01

    The effect of elongation on minimum-cost devices is investigated for elongations ranging from 2 to 3. The analysis, carried out with the TETRA tokamak systems code, includes the effects of elongation on both physics (plasma beta limit) and engineering (poloidal field coil currents) issues. When ignition is required, the minimum cost occurs for elongations from 2.3 to 2.9, depending on the plasma energy confinement scaling used. Scalings that include favorable plasma current dependence and/or degradation with fusion power tend to have minimum cost at higher elongation (2.5-2.9); scalings that depend primarily on size result in lower elongation (/approximately/2.3) for minimum cost. For design concepts that include steady-state current-driven operation, minimum cost occurs at an elongation of 2.3. 12 refs., 13 figs

  8. ZORNOC: a 1 1/2-D tokamak data analysis code for studying noncircular high beta plasmas

    International Nuclear Information System (INIS)

    Zurro, B.; Wieland, R.M.; Murakami, M.; Swain, D.W.

    1980-03-01

    A new tokamak data analysis code, ZORNOC, was developed to study noncircular, high beta plasmas in the Impurity Study Experiment (ISX-B). These plasmas exhibit significant flux surface shifts and elongation in both ohmically heated and beam-heated discharges. The MHD equilibrium flux surface geometry is determined by solving the Grad-Shafranov equation based on: (1) the shape of the outermost flux surface, deduced from the magnetic loop probes; (2) a pressure profile, deduced by means of Thomson scattering data (electrons), charge exchange data (ions), and a Fokker-Planck model (fast ions); and (3) a safety factor profile, determined from the experimental data using a simple model (Z/sub eff/ = const) that is self-consistently altered while the plasma equilibrium is iterated. For beam-heated discharches the beam deposition profile is determined by means of a Monte Carlo scheme and the slowing down of the fast ions by means of an analytical solution of the Fokker-Planck equation. The code also carries out an electron power balance and calculates various confinement parameters. The code is described and examples of its operation are given

  9. Relaxation of plasma potential and poloidal flows in the boundary of tokamak plasmas

    International Nuclear Information System (INIS)

    Hron, M.; Duran, I.; Stoeckel, J.; Hidalgo, C.; Gunn, J.

    2003-01-01

    The relaxation times of plasma parameters after a sudden change of electrode voltage have been measured in the plasma boundary during polarization experiments on the CASTOR tokamak (R = 0.4 m, a = 75 mm, B t = 1 T, I p ∼ 9 kA, q a ∼ 10). The time evolution of the floating potential after the biasing voltage switch-off can be well fitted by an exponential decay with characteristic time in the range of 10 - 20 μs. The poloidal flow shows a transient behaviour with a time scale of about 10 - 30 μs. These time scales are smaller than the expected damping time based on neoclassical parallel viscosity (which is in the range of 100 νs) and atomic physics via charge exchange (in the range of 100 - 1000 νs). But, they are larger than the correlation time of plasma turbulence (about 5 μs). These findings suggest that anomalous damping rate mechanisms for radial electric fields and poloidal flows may play a role in the boundary of tokamak plasmas. (authors)

  10. Impact of magnetic perturbation fields on tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Fietz, Sina; Maraschek, Marc; Suttrop, Wolfgang; Zohm, Hartmut [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Classen, Ivo [FOM-Institute DIFFER, Nieuwegein (Netherlands); Collaboration: the ASDEX Upgrade Team

    2015-05-01

    Non-axisymmetric external magnetic perturbation (MP) fields arise in every tokamak e.g. due to not perfectly positioned external coils. Additionally many tokamaks, like ASDEX Upgrade (AUG), are equipped with a set of external coils, which produce a 3D MP field in addition to the equilibrium field. This field is used to either compensate for the intrinsic MP field or to influence MHD instabilities such as Edge Localised Modes (ELMs) or Neoclassical Tearing Modes (NTMs). But these MP fields can also give rise to a more global plasma response. The resonant components can penetrate the plasma and influence the stability of existing NTMs or even lead to their formation via magnetic reconnection. In addition they exert a local torque on the plasma. These effects are less pronounced at high plasma rotation where the resonant field components are screened. The non-resonant components do not influence NTMs directly but slow down the plasma rotation globally via the neoclassical toroidal viscous torque. The island formation caused by the MP field as well as the interaction of pre-existing islands with the MP field at AUG is presented. It is shown that these effects can be modelled using a simple forced reconnection theory. Also the effect of resonant and non-resonant MPs on the plasma rotation at AUG is discussed.

  11. The development of 128 ch fast multi channel pulse height analyzer for a tokamak plasmas

    International Nuclear Information System (INIS)

    Kawashima, Hisato; Matoba, Tohru; Ogawa, Toshihide; Kawakami, Tomohide

    1985-02-01

    A high counting rate multi channel pulse height analyzer was developed and tested to measure the detailed time evolution of X-ray energy spectrun radiated from a tokamak plasmas. Main developing objects of this analyzer are as follows. 1. The maximum counting rate and the minimum time resolution are 4 Mcps and 10 ms, respectively. 2. The energy resolution has ability to distinguish the characterisitic X-ray line. 3. Computer has to be used for operating system. This fast multi channel analyzer is using to measure the Soft X-ray spectrum on JFT-2M tokamak, and is confirmed to be useful for a practical measuring system. (author)

  12. PELLET: a computer routine for modeling pellet fueling in tokamak plasmas

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Iskra, M.A.; Howe, H.C.; Attenberger, S.E.

    1979-01-01

    Recent experimental results of frozen hydrogenic pellet injection into hot tokamak plasmas and substantial agreement with theoretical predictions have led to a much greater interest in pellets as a means of refueling plasmas. The computer routine PELLET has been developed and used as an aid in assessing pellet ablation models and the effects of pellets on plasma behavior. PELLET provides particle source profiles under various options for the ablation model and can be coupled either to a fluid transport code or to a brief routine which supplies the required input parameters

  13. Experimental investigation of turbulent transport at the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Fedorczak, N.

    2010-01-01

    This manuscript is devoted to the experimental investigation of particle transport in the edge region of the tokamak Tore Supra. The first part introduces the motivations linked to energy production, the principle of a magnetic confinement and the elements of physics essential to describe the dynamic of the plasma at the edge region. From data collected by a set of Langmuir probes and a fast visible imaging camera, we demonstrate that the particle transport is dominated by the convection of plasma filaments, structures elongated along magnetic field lines. They present a finite wave number, responsible for the high enhancement of the particle flux at the low field side of the tokamak. This leads to the generation of strong parallel flows, and the strong constraint of filament geometry by the magnetic shear. (author)

  14. Modelling dust transport in tokamaks

    International Nuclear Information System (INIS)

    Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

    2008-01-01

    The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

  15. Magnetic analysis of tokamak plasma with approximate MHD equilibrium solution

    International Nuclear Information System (INIS)

    Moriyama, Shin-ichi; Hiraki, Naoji

    1993-01-01

    A magnetic analysis method for determining equilibrium configuration parameters (plasma shape, poloidal beta and internal inductance) on a non-circular tokamak is described. The feature is to utilize an approximate MHD equilibrium solution which explicitly relates the configuration parameters with the magnetic fields picked up by magnetic sensors. So this method is suitable for the real-time analysis performed during a tokamak discharge. A least-squares fitting procedure is added to the analytical algorithm in order to reduce the errors in the magnetic analysis. The validity is investigated through the numerical calculation for a tokamak equilibrium model. (author)

  16. Modeling of the equilibrium of a tokamak plasma

    International Nuclear Information System (INIS)

    Grandgirard, V.

    1999-12-01

    The simulation and the control of a plasma discharge in a tokamak require an efficient and accurate solving of the equilibrium because this equilibrium needs to be calculated again every microsecond to simulate discharges that can last up to 1000 seconds. The purpose of this thesis is to propose numerical methods in order to calculate these equilibrium with acceptable computer time and memory size. Chapter 1 deals with hydrodynamics equation and sets up the problem. Chapter 2 gives a method to take into account the boundary conditions. Chapter 3 is dedicated to the optimization of the inversion of the system matrix. This matrix being quasi-symmetric, the Woodbury method combined with Cholesky method has been used. This direct method has been compared with 2 iterative methods: GMRES (generalized minimal residual) and BCG (bi-conjugate gradient). The 2 last chapters study the control of the plasma equilibrium, this work is presented in the formalism of the optimized control of distributed systems and leads to non-linear equations of state and quadratic functionals that are solved numerically by a quadratic sequential method. This method is based on the replacement of the initial problem with a series of control problems involving linear equations of state. (A.C.)

  17. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    2000-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge MHD instabilities and plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. The article examines these phenomena and their interaction. (author)

  18. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    1999-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

  19. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    2001-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

  20. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-09-01

    Microinstabilities in linear and nonlinear tokamak plasmas were studied. A variational method based on the existence of a system of angular variables and action for the charged particles in the magnetic configuration of a tokamak is described. The corresponding functional, extremal in relation to the fluctuating electromagnetic field, is calculated analytically, taking into account the effects of the toroidal geometry. A numerical code, TORRID, was derived from these principles and the main instabilities, especially ion instabilities and microtearing, were studied linearly. Nonlinear methods were also applied to microtearing. Quasi-linear transport coefficients are derived from a principle of minimum entropy production. Thermal ionic conductivity and viscosity are calculated for an ionic turbulence [fr

  1. Tokamak plasma boundary layer model

    International Nuclear Information System (INIS)

    Volkov, T.F.; Kirillov, V.D.

    1983-01-01

    A model has been developed for the limiter layer and for the boundary region of the plasma column in a tokamak to facilitate analytic calculations of the thickness of the limiter layers, the profiles and boundary values of the temperature and the density under various conditions, and the difference between the electron and ion temperatures. This model can also be used to analyze the recycling of neutrals, the energy and particle losses to the wall and the limiter, and other characteristics

  2. Plasma rotation under a driven radial current in a tokamak

    International Nuclear Information System (INIS)

    Chang, C.S.

    1999-01-01

    The neoclassical behaviour of plasma rotation under a driven radial electrical current is studied in a tokamak geometry. An ambipolar radial electric field develops instantly in such a way that the driven current is balanced by a return current j p in the plasma. The j p x B torque pushes the plasma into a new rotation state both toroidally and poloidally. An anomalous toroidal viscosity is needed to avoid an extreme toroidal rotation speed. It is shown that the poloidal rotation relaxes to a new equilibrium speed, which is in general smaller than the E x B poloidal speed, and that the timescale for the relaxation of poloidal rotation is the same as that of toroidal rotation generation, which is usually much longer than the ion-ion collision time. (author)

  3. Interaction of candidate plasma facing materials with tokamak plasma in COMPASS

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Macková, Anna; Malinský, Petr; Havránek, Vladimír; Naydenkova, Diana; Klevarová, Veronika; Petersson, P.; Gasior, P.; Hakola, A.; Rubel, M.; Fortuna, E.; Kolehmainen, J.; Tervakangas, S.

    2017-01-01

    Roč. 493, September (2017), s. 102-119 ISSN 0022-3115. [International Conference on Plasma-Facing Materials and Components for Fusion Applications/15./. Aix-en-Provence, 18.05.2015-22.05.2015] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045; GA MŠk LM2015056 Institutional support: RVO:61389021 ; RVO:61389005 Keywords : erosion * COMPASS tokamak * plasma-material interaction * ion beam analysis Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (UJF-V) OBOR OECD: Nuclear related engineering ; Nuclear related engineering (UJF-V) Impact factor: 2.048, year: 2016 http://www.sciencedirect.com/science/ article /pii/S0022311517301708

  4. Heating of plasmas in tokamaks by current-driven turbulence

    International Nuclear Information System (INIS)

    Kluiver, H. de.

    1985-10-01

    Investigations of current-driven turbulence have shown the potential to heat plasmas to elevated temperatures in relatively small cross-section devices. The fundamental processes are rather well understood theoretically. Even as it is shown to be possible to relax the technical requirements on the necessary electric field and the pulse length to acceptable values, the effect of energy generation near the plasma edge, the energy transport, the impurity influx and the variation of the current profile are still unknown for present-day large-radius tokamaks. Heating of plasmas by quasi-stationary weakly turbulent states caused by moderate increases of the resistivity due to higher loop voltages could be envisaged. Power supplies able to furnish power levels 5-10 times higher than the usual values could be used for a demonstration of those regimes. At several institutes and university laboratories the study of turbulent heating in larger tokamaks and stellarators is pursued

  5. Stability of tearing modes in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1994-02-01

    The stability properties of m ≥ 2 tearing instabilities in tokamak plasmas are analyzed. A boundary layer theory is used to find asymptotic solutions to the ideal external kink equation which are used to obtain a simple analytic expression for the tearing instability parameter Δ'. This calculation generalizes previous work on this topic by considering more general toroidal equilibria (however, toroidal coupling effects are ignored). Constructions of Δ' are obtained for plasmas with finite beta and for islands that have nonzero width. A simple heuristic estimate is given for the value of the saturated island width when the instability criterion is violated. A connection is made between the calculation of the asymptotic matching parameter in the finite beta and island width case to the nonlinear analog of the Glasser effect

  6. Modeling of the equilibrium of a tokamak plasma; Modelisation de l'equilibre d'un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Grandgirard, V

    1999-12-01

    The simulation and the control of a plasma discharge in a tokamak require an efficient and accurate solving of the equilibrium because this equilibrium needs to be calculated again every microsecond to simulate discharges that can last up to 1000 seconds. The purpose of this thesis is to propose numerical methods in order to calculate these equilibrium with acceptable computer time and memory size. Chapter 1 deals with hydrodynamics equation and sets up the problem. Chapter 2 gives a method to take into account the boundary conditions. Chapter 3 is dedicated to the optimization of the inversion of the system matrix. This matrix being quasi-symmetric, the Woodbury method combined with Cholesky method has been used. This direct method has been compared with 2 iterative methods: GMRES (generalized minimal residual) and BCG (bi-conjugate gradient). The 2 last chapters study the control of the plasma equilibrium, this work is presented in the formalism of the optimized control of distributed systems and leads to non-linear equations of state and quadratic functionals that are solved numerically by a quadratic sequential method. This method is based on the replacement of the initial problem with a series of control problems involving linear equations of state. (A.C.)

  7. 3D simulation studies of tokamak plasmas using MHD and extended-MHD models

    International Nuclear Information System (INIS)

    Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.

    1996-01-01

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

  8. Tokamak-7 operation in experiments with a plasma

    International Nuclear Information System (INIS)

    Buzanki, V.V.; Bychkov, A.V.; Denisov, V.F.

    1982-01-01

    The results of experiments with plasma at the Tokamak-7 (T-7) device are presented. The experiments have been carried out with a constant diaphragm of 31,5 cm radius and two movable graphite diaphragms at the 26-28 cm plasma filament radius and 1,6-1,9 T magnetic field. Two stable regimes with 150 and 200 kA and 250 ms discharge current length have been investigated. It is shown that the strongest poloidal filed perturhations have been observed at the beginning of the discharge. Electron plasma temperature Tsub(e) has been determined from the spectrum analysis of soft X radiation by the foil method. Stable plasma regimes with current up to 200 kA, bypass voltage being equal 1,58V electron density -0,5-5,0 x 10 13 cm -3 , Tsub(e)=1,1-1,3 keV ion temperature-490 eV. The range between discharge pulses has reached 3 min. at the discharge current-240 kA. No considerable effect of magnetic field variables on the superconducting magnetic system has been observed

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

    International Nuclear Information System (INIS)

    Peters, M.

    1996-01-01

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

  10. 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.).

  11. Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta

    International Nuclear Information System (INIS)

    Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.

    2017-01-01

    Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t ), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.

  12. Modeling plasma/material interactions during a tokamak disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1994-10-01

    Disruptions in tokamak reactors are still of serious concern and present a potential obstacle for successful operation and reliable design. Erosion of plasma-facing materials due to thermal energy dump during a disruption can severely limit the lifetime of these components, therefore diminishing the economic feasibility of the reactor. A comprehensive disruption erosion model which takes into account the interplay of major physical processes during plasma-material interaction has been developed. The initial burst of energy delivered to facing-material surfaces from direct impact of plasma particles causes sudden ablation of these materials. As a result, a vapor cloud is formed in front of the incident plasma particles. Shortly thereafter, the plasma particles are stopped in the vapor cloud, heating and ionizing it. The energy transmitted to the material surfaces is then dominated by photon radiation. It is the dynamics and the evolution of this vapor cloud that finally determines the net erosion rate and, consequently, the component lifetime. The model integrates with sufficient detail and in a self-consistent way, material thermal evolution response, plasma-vapor interaction physics, vapor hydrodynamics, and radiation transport in order to realistically simulate the effects of a plasma disruption on plasma-facing components. Candidate materials such as beryllium and carbon have been analyzed. The dependence of the net erosion rate on disruption physics and various parameters was analyzed and is discussed

  13. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, P.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.K.; Owen, L.; Matthews, G.

    1990-01-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p T < or approx.10.7%, P(auxiliary)< or approx.20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions. (orig.)

  14. The critical temperature gradient model of plasma transport: applications to Jet and future tokamaks

    International Nuclear Information System (INIS)

    Rebut, P.H.; Lallia, P.P.; Watkins, M.L.

    1989-01-01

    The diversity and complexity of behaviour in tokamak plasmas place strong constraints on any model attempting a description in terms of a single underlying phenomenon. Assuming that turbulence in the magnetic topology is the underlying phenomenon, specific expressions for electron and ion heat flux are derived from heuristic and dimensional arguments. When used in plasma transport codes, rather satisfactory simulations of experimental results are achieved in different sized tokamaks in various regimes of operation. Predictions are given for the expected performance of JET at full planned power and implications for next step tokamaks are indicated

  15. Reactor aspects of counterstreaming-ion tokamak plasmas

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1975-06-01

    Toroidal DT plasmas in which the D and T ions make up two distinct, quasi-thermal velocity distributions, oppositely displaced in velocity along the magnetic axis, are discussed. Such counterstreaming distributions can be set up by introducing all ions by tangential injection of neutral beams, and by removing ions from the plasma shortly after they have decelerated to an energy approximate to or less than 2T/sub e/ by Coulomb drag on the plasma electrons. A simple physical model for counterstreaming-ion operation is postulated, which allows one to deduce the ion velocity distributions and required energy and particle confinement times that are in good agreement with the results of previous Fokker-Planck calculations. The variations of fusion reactivity, power gain, and power density with injection energy and electron temperature are presented. The practical problems of implementing counter-streaming operation in a tokamak, such as charge-exchange losses, the prompt removal of cold ions, and the effect of impurities are discussed. (U.S.)

  16. New DIII-D tokamak plasma control system

    International Nuclear Information System (INIS)

    Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.

    1992-09-01

    A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter

  17. Resistive internal kink modes in a tokamak with high-pressure plasma

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Mikhajlovskij, A.B.; Tatarinov, E.G.

    1988-01-01

    Theory of resistive internal kink modes in a tokamak with high-pressure plasma is developed. Equation for Fourie-image of disturbed displacment in a resistive layer ie derived with regard to effects of the fourth order by plasma pressure within the framework of single-liquid approach. In its structure this equation coincides with a similar equation for resistive balloon modes and has an exact solution expressed by degenerated hypergeometric function. A general dispersion equation for resistive kink modes is derived with regard to the effects indicated. It is shown that plasma pressure finiteness leads to the reduction of reconnection and tyring-mode increments

  18. Soft x-ray imaging system for measurement of noncircular tokamak plasmas

    International Nuclear Information System (INIS)

    Fonck, R.J.; Reusch, M.; Jaehnig, K.P.; Hulse, R.; Roney, P.

    1986-08-01

    A soft x-ray camera and image processing system has been constructed to provide measurements of the internal shape of high temperature tokamak plasmas. The camera consists of a metallic-foil-filtered pinhole aperture and a microchannel plate image intensifier/convertor which produces a visible image for detection by a CCD TV camera. A wide-angle tangential view of the toroidal plasma allows a single compact camera to view the entire plasma cross section. With Be filters 12 to 50 μm thick, the signal from the microchannel plate is produced mostly by nickel L-line emissions which orignate in the hot plasma core. The measured toroidal image is numerically inverted to produce a cross-sectional soft x-ray image of the plasma. Since the internal magnetic flux surfaces are usually isothermal and the nickel emissivity depends strongly on the local electron temperature, the x-ray emission contours reflect the shape of the magnetic surfaces in the plasma interior. Initial results from the PBX tokamak experiment show clear differences in internal plasma shapes for circular and bean-shaped discharges

  19. Relaxation of potential, flows, and density in the edge plasma of Castor tokamak

    International Nuclear Information System (INIS)

    Hron, M.; Weinzettl, V.; Dufkova, E.; Duran, I.; Stoeckel, J.; Hidalgo, C.

    2004-01-01

    Decay times of plasma flows and plasma profiles have been measured after a sudden biasing switch-off in experiments on the Castor tokamak. A biased electrode has been used to polarize the edge plasma. The edge plasma potential and flows have been characterized by means of Langmuir and Mach probes, the radiation was measured using an array of bolometers. Potential profiles and poloidal flows can be well fitted by an exponential decay time in the range of 10 - 30 μs when the electrode biasing is turned off in the Castor tokamak. The radiation shows a slower time scale (about 1 ms), which is linked to the evolution in the plasma density and particle confinement. (authors)

  20. Development of anode high voltage power supply system for ECRH of HL-2A tokamak

    International Nuclear Information System (INIS)

    Chen Wenguang

    2009-01-01

    The anode high voltage power supply system consist of DC high-voltage power supply (HVPS) and pulse modulator. SCR is used to vary AC input voltage of the step-up transformer by controlling the trigger phase in the HVPS, and regulate the DC output voltage linearly at the potential of low-end via BJT, Dual closed-loop control technology is applied in the controller, and its maximum output is at 30kV and 130mA. Tetrode is the core component of the modulator. The circuit design is optimized by using the simulation software. Test and HL-2A discharge experimental results show that the power supply system is designed with some characteristics of output scale widely, low ripple and modulate quickly. (authors)

  1. Equilibrium of rotating and nonrotating plasmas in tokamaks

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2003-01-01

    One studied plasma equilibrium in tokamak in case of toroidal rotation. Rotation associated centrifugal force is shown to result in decrease of equilibrium limit as to β. One analyzes unlike opinion and considers its supports. It is shown that in possible case of local improvement of equilibrium conditions associated with special selection of profile of plasma rotation rate, the combined integral effect turns to be negative one. But in case of typical conditions, decrease of equilibrium β caused by plasma rotation is negligible one and one may ignore effect of plasma rotation on its equilibrium for hot plasma [ru

  2. The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

    International Nuclear Information System (INIS)

    1993-12-01

    The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support

  3. Determination of the plasma position for its real-time control in the COMPASS tokamak

    International Nuclear Information System (INIS)

    Janky, F.; Havlicek, J.; Valcarcel, D.; Hron, M.; Horacek, J.; Kudlacek, O.; Panek, R.; Carvalho, B.B.

    2011-01-01

    An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

  4. Determination of the plasma position for its real-time control in the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Havlicek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Valcarcel, D. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal); Hron, M.; Horacek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Kudlacek, O. [Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, Technicka 2, 166 27 Prague (Czech Republic); Panek, R. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Carvalho, B.B. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal)

    2011-10-15

    An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

  5. Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Chrystal, C. [University of California-San Diego, La Jolla, California 92186-5608 (United States); Burrell, K. H.; Groebner, R. J.; Kaplan, D. H. [General Atomics, San Diego, California 92186-5608 (United States); Grierson, B. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)

    2012-10-15

    To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

  6. Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma.

    Science.gov (United States)

    Chrystal, C; Burrell, K H; Grierson, B A; Groebner, R J; Kaplan, D H

    2012-10-01

    To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

  7. Radio-frequency heating and neutral atom transport in a fluid-magnetohydrodynamic treatment of burning tokamak plasmas

    International Nuclear Information System (INIS)

    Conn, R.W.; Mau, T.K.; Prinja, A.K.

    1983-01-01

    A physical model for the space and time evolution of the primary parameters of ordinary and burning tokamak plasmas is described by employing a fluid plasma treatment coupled to a magnetohydrodynamic equilibrium description, the solution to the appropriate Maxwell equations, and the solution of the linear transport equation describing neutral atom transport in plasmas. The specific problems of plasma heating by ion cyclotron radiofrequency (ICRF) waves and neutral atom transport in the plasma edge and in complicated geometrical components such as divertor channels or pumped limiter structures are analyzed. A theoretical, onedimensional slab model of ICRF heating at ω = 2ω/SUB cD/ is developed and applied to determine the space-time response of tokamak plasmas. Generally, strong single-pass absorption is found for high-density, high (β) plasmas using a low k 11 spectrum (0.05 to 0.1 cm -1 ) although for (β > 1%, electron Landau damping becomes important. Deterministic and Monte Carlo methods to solve the neutral atom transport problem are described. Specific application to determine the spectrum of neutral atoms emerging from the duct of a pump limiter shows it to be hard (mean energy > 20 eV), indicating very incomplete energy thermalization. Uncertainties are identified in the overall problem of dynamic burning plasma analysis caused by the complexity of the problem itself and by uncertainties in fundamental areas such as plasma transport coefficients, stability, and plasma edge physics

  8. Understanding of hysteresis behaviors at the L-H-L transitions in tokamak plasma based on bifurcation concept

    Energy Technology Data Exchange (ETDEWEB)

    Chatthong, B. [Department of Physics, Faculty of Science, Prince of Songkla University, Hat Yai, Songkla (Thailand); Onjun, T. [School of Manufacturing Systems and Mechanical Engineering, Sirindhorn International Institute of Technology, Thammasat University, Pathum Thani (Thailand)

    2016-08-15

    The hysteresis behaviour at the L-H-L transitions in tokamak plasma is investigated based on bifurcation concept. The formation of an edge transport barrier (ETB) is modeled via thermal and particle transport equations with the flow shear suppression effect on anomalous transport included. The anomalous transport is modeled based on critical gradients threshold and the flow shear is calculated from the force balance equation, couples the two transport equations leading to a non-linear behaviour. Analytical investigation reveals that the fluxes versus gradients space exhibits bifurcation behaviour with s -curve soft bifurcation type. Apparently, the backward H-L transition occurs at lower values than that of the forward L-H transition, illustrating hysteresis behaviour. The hysteresis properties, i.e. locations of threshold fluxes, gradients and their ratios are analyzed as a function of neoclassical and anomalous transport values and critical gradients. It is found that the minimum heat flux for maintaining H -mode depends on several plasma parameters including the strength of anomalous transport and neoclassical transport. In particular, the hysteresis depth becomes larger when neoclassical transport decreases or anomalous transport increases. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  9. Feedback control of horizontal position and plasma surface shape in a non-circular tokamak

    International Nuclear Information System (INIS)

    Moriyama, Shin-ichi; Nakamura, Kazuo; Nakamura, Yukio; Itoh, Satoshi

    1986-01-01

    The linear model for the coupled horizontal position and plasma surface shape control in the non-circular tokamak device was described. It enables us to estimate easily the displacement and the distortion due to the changes in plasma pressure and current density distribution. The PI-controller and the optimal regulator were designed with the linear model. Transient-response analysis of the control system in the TRIAM-1M tokamak showed that the optimal regulator is superior to the PI-controller with regard to the mutual-interference between the position control system and the elongation control system. (author)

  10. Plasma rotation measurement in small tokamaks using an optical spectrometer and a single photomultiplier as detector.

    Science.gov (United States)

    Severo, J H F; Nascimento, I C; Kuznetov, Yu K; Tsypin, V S; Galvão, R M O; Tendler, M

    2007-04-01

    The method for plasma rotation measurement in the tokamak TCABR is reported in this article. During a discharge, an optical spectrometer is used to scan sequentially spectral lines of plasma impurities and spectral lines of a calibration lamp. Knowing the scanning velocity of the diffraction grating of the spectrometer with adequate precision, the Doppler shifts of impurity lines are determined. The photomultiplier output voltage signals are recorded with adequate sampling rate. With this method the residual poloidal and toroidal plasma rotation velocities were determined, assuming that they are the same as those of the impurity ions. The results show reasonable agreement with the neoclassical theory and with results from similar tokamaks.

  11. Plasma fluctuation measurements in tokamaks using beam-plasma interactions

    International Nuclear Information System (INIS)

    Fonck, R.J.; Duperrex, P.A.; Paul, S.F.

    1990-01-01

    High-frequency observations of light emitted from the interactions between plasma ions and injected neutral beam atoms allow the measurement of moderate-wavelength fluctuations in plasma and impurity ion densities. To detect turbulence in the local plasma ion density, the collisionally excited fluorescence from a neutral beam is measured either separately at several spatial points or with a multichannel imaging detector. Similarly, the role of impurity ion density fluctuations is measured using charge exchange recombination excited transitions emitted by the ion species of interest. This technique can access the relatively unexplored region of long-wavelength plasma turbulence with k perpendicular ρ i much-lt 1, and hence complements measurements from scattering experiments. Optimization of neutral beam geometry and optical sightlines can result in very good localization and resolution (Δx≤1 cm) in the hot plasma core region. The detectable fluctuation level is determined by photon statistics, atomic excitation processes, and beam stability, but can be as low as 0.2% in a 100 kHz bandwidth over the 0--1 MHz frequency range. The choices of beam species (e.g., H 0 , He 0 , etc.), observed transition (e.g., H α , L α , He I singlet or triplet transitions, C VI Δn=1, etc.) are dictated by experiment-specific factors such as optical access, flexibility of beam operation, plasma conditions, and detailed experimental goals. Initial tests on the PBX-M tokamak using the H α emissions from a heating neutral beam show low-frequency turbulence in the edge plasma region

  12. Non-inductive plasma initiation and plasma current ramp-up on the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Takase, Y.; Ejiri, A.; Oosako, T.; Shinya, T.; Ambo, T.; Furui, H.; Kato, K.; Nakanishi, A.; Sakamoto, T.; Kakuda, H.; Wakatsuki, T.; Hashimoto, T.; Hiratsuka, J.; Kasahara, H.; Kumazawa, R.; Mutoh, T.; Saito, K.; Seki, T.; Moeller, C.P.; Nagashima, Y.

    2013-01-01

    Plasma current (I p ) start-up in a spherical tokamak (ST) by waves in the lower-hybrid (LH) frequency range was investigated on TST-2. A low current (∼1 kA) ST configuration can be formed by waves over a broad frequency range (21 MHz–8.2 GHz in TST-2), but further I p ramp-up (to ∼10 kA) is most efficient with waves in the LH frequency range. I p ramp-up to 15 kA was achieved with 60 kW of net RF power P RF in the fast wave (FW) polarization at 200 MHz excited by the inductively coupled combline antenna. X-ray measurements showed that the photon flux and temperature are higher in the direction opposite to I p , consistent with acceleration of electrons by a uni-directional RF wave. There is evidence that the LH wave is excited nonlinearly by the FW, based on the frequency spectra measured by magnetic probes. Similar efficiencies of I p ramp-up were obtained with the inductive combline antenna and the dielectric-loaded waveguide array (‘grill’) antenna, and tendencies for the current drive efficiency to increase with plasma current and toroidal field were observed. During operation of the grill antenna, wavevector components were measured by an array of magnetic probes. Results were qualitatively consistent with expectations based on dispersion relations for the FW and the LH wave. A capacitively coupled combline antenna has been developed to improve coupling to the plasma and the wavenumber spectrum of the excited LH wave, and will be tested in 2013. (paper)

  13. Neoclassical Physics for Current Drive in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Duthoit, F.X.

    2012-03-01

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

  14. Spectra of germanium and selenium in the 50-350 A region from the PLT tokamak plasma

    International Nuclear Information System (INIS)

    Stratton, B.C.; Hodge, W.L.; Moos, H.W.; Schwob, J.L.; Suckewer, S.; Finkenthal, M.; Cohen, S.

    1983-03-01

    Spectra of germanium and selenium injected into the PLT tokamak plasma were observed in the 50 to 350 A region for GeXIV-XXV (KI to OI-like) and SeXVI-XXIV (KI to NaI-like). A number of 3p/sup k/-3p/sup k-1/3d transitions predicted by isoelectronic sequence extrapolation have been identified. Also, previously identified lines from ions in the AlI to OI-like and KI-like isoelectronic sequences have been observed in the tokamak plasma

  15. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe

    Czech Academy of Sciences Publication Activity Database

    Walkden, N.R.; Adámek, Jiří; Allan, S.; Dudson, B.D.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Komm, Michael

    2015-01-01

    Roč. 86, č. 2 (2015), č. článku 023510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * ball pen probe Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.336, year: 2015 http://dx.doi.org/10.1063/1.4908572

  16. Designing a Sine-Coil for Measurement of Plasma Displacements in IR-T1 Tokamak

    International Nuclear Information System (INIS)

    Khorshid, Pejman; Razavi, M.; Molaii, M.; Ghoranneviss, M.; TalebiTaher, A.; Arvin, R.; Mohammadi, S.; NikMohammadi, A.

    2008-01-01

    A method for the measurement of the plasma position in the IR-T1 tokamak in toroidal coordinates is developed. A sine-coil, which is a Rogowski coil with a variable wiring density is designed and fabricated for this purpose. An analytic solution of the Biot-Savart law, which is used to calculate magnetic fields created by toroidal plasma current, is presented. Results of calculations are compared with the experimental data obtained in no-plasma shots with a toroidal current-carrying coil positioned inside the vessel to simulate the plasma movements. The results are shown a good linear behavior of plasma position measurements. The error is less than 2.5% and it is compared with other methods of measurements of the plasma position. This method will be used in the feedback position control system and tests of feedback controller parameters are ongoing

  17. Computation of electromagnetic effects in a tokamak due to a plasma disruption

    International Nuclear Information System (INIS)

    Turner, L.R.

    1989-01-01

    To model the consequences of a plasma disruption in a tokamak one must combine a code that computes the detailed MHD behavior of the plasma with one that treats the three-dimensional features of the conducting toroidal components around the plasma. The NET (Next European Torus) Team have undertaken a treatment of electromagnetic effects from plasma disruptions using both open loop and closed loop integration of codes. In America, workers at Oak Ridge National Laboratory, Idaho National Engineering Laboratory, and Argonne National Laboratory have looked at plasma disruption effects on the ITER blanket using the codes TSC and EDDYNET. Results show how the forces on a blanket segment depend on the number and size of the segments and on the gap between them. 9 refs., 4 figs., 1 tab

  18. Confinement of ohmically heated plasmas and turbulent heating in high-magnetic field tokamak TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

    Hiraki, N; Itoh, S; Kawai, Y; Toi, K; Nakamura, K [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

    1979-12-01

    TRIAM-1, the tokamak device with high toroidal magnetic field, has been constructed to establish the scaling laws of advanced tokamak devices such as Alcator, and to study the possibility of the turbulent heating as a further economical heating method of the fusion oriented plasmas. The plasma parameters obtained by ohmic heating alone are as follows; central electron temperature T sub(e0) = 640 eV, central ion temperature T sub(i0) = 280 eV and line-average electron density n average sub(e) = 2.2 x 10/sup 14/ cm/sup -3/. The empirical scaling laws are investigated concerning T sub(e0), T sub(i0) and n average sub(e). The turbulent heating has been carried out by applying the high electric field in the toroidal direction to the typical tokamak discharge with T sub(i0) asymptotically equals 200 eV. The efficient ion heating is observed and T sub(i0) attains to about 600 eV.

  19. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, T.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.; Owen, L.; Matthews, G.

    1990-06-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p approx-lt 3 MA, β T approx-lt 10.7%, P(auxiliary) approx-lt 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma surface interactions. 41 refs., 8 figs., 1 tab

  20. Peeling-off of the external kink modes at tokamak plasma edge

    International Nuclear Information System (INIS)

    Zheng, L. J.; Furukawa, M.

    2014-01-01

    It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture

  1. Peeling-off of the external kink modes at tokamak plasma edge

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, L. J. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States); Furukawa, M. [Graduate School of Engineering, Tottori University, Tottori 680-8552 (Japan)

    2014-08-15

    It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture.

  2. On thermionic emission from plasma-facing components in tokamak-relevant conditions.

    Czech Academy of Sciences Publication Activity Database

    Komm, Michael; Ratynskaia, S.; Tolias, P.; Cavalier, Jordan; Dejarnac, Renaud; Gunn, J. P.; Podolník, Aleš

    2017-01-01

    Roč. 59, č. 9 (2017), č. článku 094002. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : thermionic * PIC * tungsten * tokamak * thermionic emission * plasma facing components * particle-in-cell Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/1361-6587/aa78c4/pdf

  3. Large plasma pressure perturbations and radial convective transport in a tokamak

    International Nuclear Information System (INIS)

    Krasheninnikov, Sergei; Yu, Guanghui; Ryutov, Dmitri

    2004-01-01

    Strongly localized plasma structures with large pressure inhomogeneities (such as plasma blobs in the scrape-off-layer (SOL)/shadow regions, pellet clouds, Edge localized Modes (ELMs)) observed in the tokamaks, stellarators and linear plasma devices. Experimental studies of these phenomena reveal striking similarities including more convective rather than diffusive radial plasma transport. We suggest that rather simple models can describe many essentials of blobs, ELMs, and pellet clouds dynamics. The main ingredient of these models is the effective plasma gravity caused by magnetic curvature, centrifugal or friction forces effects. As a result, the equations governing plasma transport in such localized structures appear to be rather similar to that used to describe nonlinear evolution of thermal convection in the Boussinesq approximation (directly related to the Rayleigh-Taylor (RT) instability). (author)

  4. Edge Plasma Physics and Relevant Diagnostics on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Devynck, P.; Gunn, J.; Martines, E.; Bonhomme, G.; Van Oost, G.; Hron, Martin; Ďuran, Ivan; Pánek, Radomír; Stejskal, Pavel; Adámek, Jiří

    2004-01-01

    Roč. 3, - (2004), s. 1-6 ISSN 1433-5581. [First Cairo Conference on Plasma Physics & Applications. Cairo, 11.10.2003-15.10.2003] R&D Projects: GA ČR GA202/03/0786; GA ČR GP202/03/P062 Keywords : tokamak * edge plasma * probe diagnostics * biasing * turbulence * polarization Subject RIV: BL - Plasma and Gas Discharge Physics

  5. Advanced probes for edge plasma diagnostics on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Adámek, Jiří; Balan, P.; Hronová-Bilyková, Olena; Brotánková, Jana; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Horáček, Jan; Ionita, C.; Kocan, M.; Martines, E.; Pánek, Radomír; Peleman, P.; Schrittwieser, R.; Van Oost, G.; Žáček, František

    2006-01-01

    Roč. 63, č. 0 (2006), 012001-012002 E-ISSN 1742-6596. [SECOND INTERNATIONAL WORKSHOP AND SUMMER SCHOOL ON PLASMA PHYSICS. Kiten, 03.07.2006-09.07.2006] R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma * tokamak * electric probes * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  6. Neutral beam in ALVAND IIC tokamak

    International Nuclear Information System (INIS)

    Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

    1992-01-01

    Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

  7. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

    also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...

  8. Tokamak plasma equilibrium problems with anisotropic pressure and rotation and their numerical solution

    International Nuclear Information System (INIS)

    Ivanov, A. A.; Martynov, A. A.; Medvedev, S. Yu.; Poshekhonov, Yu. Yu.

    2015-01-01

    In the MHD tokamak plasma theory, the plasma pressure is usually assumed to be isotropic. However, plasma heating by neutral beam injection and RF heating can lead to a strong anisotropy of plasma parameters and rotation of the plasma. The development of MHD equilibrium theory taking into account the plasma inertia and anisotropic pressure began a long time ago, but until now it has not been consistently applied in computational codes for engineering calculations of the plasma equilibrium and evolution in tokamak. This paper contains a detailed derivation of the axisymmetric plasma equilibrium equation in the most general form (with arbitrary rotation and anisotropic pressure) and description of the specialized version of the SPIDER code. The original method of calculation of the equilibrium with an anisotropic pressure and a prescribed rotational transform profile is proposed. Examples of calculations and discussion of the results are also presented

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

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1977-01-01

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

  10. Transport and stability studies in negative central shear advanced tokamak plasmas

    International Nuclear Information System (INIS)

    Jayakumar, R.J.

    2003-01-01

    Achieving high performance for long duration is a key goal of Advanced Tokamak (AT) research around the world. To this end, tokamak experiments are focusing on obtaining (a) a high fraction of well-aligned non-inductive plasma current (b) wide internal transport barriers (ITBs) in the ion and electron transport channels to obtain high temperatures (c) control of resistive wall modes and neoclassical Tearing Modes which limit the achievable beta. A current profile that yields a negative central magnetic shear (NCS) in the core is consistent with the above focus; Negative central shear is conducive for obtaining internal transport barriers, for high degree of bootstrap current alignment and for reaching the second stability region for ideal ballooning modes, while being stable to ideal kink modes at high beta with wall stabilization. Much progress has been made in obtaining AT performance in several tokamaks through an increasing understanding of the stability and transport properties of tokamak plasmas. RF and neutral beam current drive scenarios are routinely developed and implemented in experiments to access new advanced regimes and control plasma profiles. Short duration and sustained Internal Transport Barriers (ITB) have been obtained in the ion and electron channels. The formation of an ITB is attributable to the stabilization of ion and electron temperature gradient (ITG and ETG) and trapped electron modes (TEM), enhancement of E x B flow shear rate and rarefaction of resonant surfaces near the rational q min values. (orig.)

  11. Transport and stability studies in negative central shear advanced tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, R.J. [Lawrence Livermore National Laboratory (United States)

    2003-07-01

    Achieving high performance for long duration is a key goal of Advanced Tokamak (AT) research around the world. To this end, tokamak experiments are focusing on obtaining (a) a high fraction of well-aligned non-inductive plasma current (b) wide internal transport barriers (ITBs) in the ion and electron transport channels to obtain high temperatures (c) control of resistive wall modes and neoclassical Tearing Modes which limit the achievable beta. A current profile that yields a negative central magnetic shear (NCS) in the core is consistent with the above focus; Negative central shear is conducive for obtaining internal transport barriers, for high degree of bootstrap current alignment and for reaching the second stability region for ideal ballooning modes, while being stable to ideal kink modes at high beta with wall stabilization. Much progress has been made in obtaining AT performance in several tokamaks through an increasing understanding of the stability and transport properties of tokamak plasmas. RF and neutral beam current drive scenarios are routinely developed and implemented in experiments to access new advanced regimes and control plasma profiles. Short duration and sustained Internal Transport Barriers (ITB) have been obtained in the ion and electron channels. The formation of an ITB is attributable to the stabilization of ion and electron temperature gradient (ITG and ETG) and trapped electron modes (TEM), enhancement of E x B flow shear rate and rarefaction of resonant surfaces near the rational q{sub min} values. (orig.)

  12. Sustainment of spherical tokamak by means of repetitive injection of compact torus plasma

    International Nuclear Information System (INIS)

    Shimamura, Shin; Matsura, Ken; Takahashi, Tsutomu; Nogi, Yasuyuki

    2000-01-01

    Sustainment of spherical tokamak (S.T.) has been studied. A compact torus (C.T.) plasma was injected into confinement region by magnetized coaxial gun. For start-up and sustainment of large main spherical tokamak, single pulsed injection of small C.T. is not sufficient in many cases. C.T.plasma injection of high repetition rate is required. For this purpose magnetized coaxial gun was driven with high repetition rate current. The first injected C.T. plasma could start-up S.T. without other help. The repetitive C.T. injection grew and sustained the S.T. plasma. A CCD camera with fast gated image intensifier took a cross sectional view of S.T. during the repetitive C.T. injection. (author)

  13. Real-time control of current and pressure profiles in tokamak plasmas

    International Nuclear Information System (INIS)

    Laborde, L.

    2005-12-01

    Recent progress in the field of 'advanced tokamak scenarios' prefigure the operation regime of a future thermonuclear fusion power plant. Compared to the reference regime, these scenarios offer a longer plasma confinement time thanks to increased magnetohydrodynamic stability and to a better particle and energy confinement through a reduction of plasma turbulence. This should give access to comparable fusion performances at reduced plasma current and could lead to a steady state fusion reactor since the plasma current could be entirely generated non-inductively. Access to this kind of regime is provided by the existence of an internal transport barrier, linked to the current profile evolution in the plasma, which leads to steep temperature and pressure profiles. The comparison between heat transport simulations and experiments allowed the nature of the barriers to be better understood as a region of strongly reduced turbulence. Thus, the control of this barrier in a stationary manner would be a remarkable progress, in particular in view of the experimental reactor ITER. The Tore Supra and JET tokamaks, based in France and in the United Kingdom, constitute ideal instruments for such experiments: the first one allows stationary plasmas to be maintained during several minutes whereas the second one provides unique fusion performances. In Tore Supra, real-time control experiments have been accomplished where the current profile width and the pressure profile gradient were controlled in a stationary manner using heating and current drive systems as actuators. In the JET tokamak, the determination of an empirical static model of the plasma allowed the current and pressure profiles to be simultaneously controlled and so an internal transport barrier to be sustained. Finally, the identification of a dynamic model of the plasma led to the definition of a new controller capable, in principle, of a more efficient control. (author)

  14. Current density distribution during disruptions and sawteeth in a simple model of plasma current in a tokamak

    International Nuclear Information System (INIS)

    Stefanovskii, A. M.

    2011-01-01

    The processes that are likely to accompany discharge disruptions and sawteeth in a tokamak are considered in a simple plasma current model. The redistribution of the current density in plasma is supposed to be primarily governed by the onset of the MHD-instability-driven turbulent plasma mixing in a finite region of the current column. For different disruption conditions, the variation in the total plasma current (the appearance of a characteristic spike) is also calculated. It is found that the numerical shape and amplitude of the total current spikes during disruptions approximately coincide with those measured in some tokamak experiments. Under the assumptions adopted in the model, the physical mechanism for the formation of the spikes is determined. The mechanism is attributed to the diffusion of the negative current density at the column edge into the zero-conductivity region. The numerical current density distributions in the plasma during the sawteeth differ from the literature data.

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

    International Nuclear Information System (INIS)

    Lister, J.B.; Villard, L.; Ridder, G. de

    1997-01-01

    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

  16. A review of the methods to measure the ion temperature in a tokamak plasma

    International Nuclear Information System (INIS)

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

    1976-01-01

    The most important methods to measure the ion temperatu--re in a Tokamak plasma are reviewed, e.g. energy analysis of the fast neutrals which leave out the plasma, Doppler broadening of the emision spectral lines and fusion neutron analysis. It is discussed their bounds so as the advantages and drawbacks of each one. Other methods of some interest in the future are outlined. (author) [es

  17. Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

    Science.gov (United States)

    Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; Grierson, B. A.; Staebler, G. M.; Rice, J. E.; Yuan, X.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Sciortino, F.

    2018-02-01

    A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. This Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and time scales of cold-pulse experiments in tokamak plasmas.

  18. Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Abiteboul, J.

    2012-10-01

    The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are

  19. 3D simulation studies of tokamak plasmas using MHD and extended-MHD models

    International Nuclear Information System (INIS)

    Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.; Pomphrey, N.; Sugiyama, L.E.

    1997-01-01

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

  20. The conversion of the thermal energy of plasma in the SOL of tokamaks

    International Nuclear Information System (INIS)

    Nedospasov, A.V.; Nenova, N.V.

    2008-01-01

    When the plasma expands across the confining magnetic field, a part of its thermal energy is converted to electrical energy. In the SOL of tokamaks, the motion of the plasma across the field due to turbulent processes is accompanied by its departure along the open lines of the magnetic field. The conversion of thermal energy is taken into account in theoretical studies devoted to the physics of plasma in the SOL; however, this conversion is ignored in numerical models, for example, in B2-SOLPS4.0. This paper deals with thermal-to-electrical energy conversion in the SOL of tokamaks. It is demonstrated that the part of the thermal energy subjected to conversion to electrical energy forms an appreciable part of the total energy flowing in the SOL. In ITER, this fraction may be as high as 20-25%. The electrical energy generated in the SOL volume is liberated in the form of Joule heat in a relatively cold plasma in the vicinity of diverter plates or directly on these plates. (letter)

  1. Magnetic sensorless control of plasma position and shape in a tokamak

    International Nuclear Information System (INIS)

    Nakamura, K.; Luo, J.R.; Wang, H.Z.

    2005-01-01

    Magnetic sensorless sensing and control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The sensing is made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of control experiments were carried out, to keep the position constant and swing the position in a triangular waveform. And magnetic sensorless sensing of plasma shape is discussed. (author)

  2. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    2001-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  3. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    1999-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

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

    International Nuclear Information System (INIS)

    Suratia, Pooja; Patel, Jigneshkumar; Rajpal, Rachana; Kotia, Sorum; Govindarajan, J.

    2012-01-01

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

  5. Analysis of tokamak plasma confinement modes using the fast Fourier transformation

    International Nuclear Information System (INIS)

    Mirmoeini, S.R.; Salar Elahi, A.; Ghoranneviss, M.

    2016-01-01

    The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and without external field (electric biasing), and then compared it with each other. After the Fourier analysis of Mirnov coil data, the diagram of power spectrum density was depicted in different angles of Mirnov coils in the 'presence of external field' as well as in the 'absence of external field'. The power spectrum density (PSD) interprets the manner of power distribution of a signal with frequency. In this article, the number of plasma modes and the safety factor q were obtained by using the mode number of q = m/n (m is the mode number). The maximum MHD activity was obtained in 30-35 kHz frequency, using the density of the energy spectrum. In addition, the number of different modes across 0-35 ms time was compared with each other in the presence and absence of the external field. (author)

  6. Runaway electrons in disruptions and perturbed magnetic topologies of tokamak plasmas

    International Nuclear Information System (INIS)

    Forster, Michael

    2012-01-01

    Nuclear fusion represents a valuable perspective for a safe and reliable energy supply from the middle of the 21st century on. Currently, the tokamak is the most advanced principle of confining a man-made fusion plasma. The operation of future, reactor sized tokamaks like ITER faces a crucial difficulty in the generation of runaway electrons. The runaway of electrons is a free fall acceleration into the relativistic regime which is known in various kinds of plasmas including astrophysical ones, thunderbolts and fusion plasmas. The tokamak disruption instability can include the conversion of a substantial part of the plasma current into a runaway electron current. When the high energetic runaways are lost, they can strike the plasma facing components at localised spots. Due to their high energies up to a few tens of MeV, the runaways carry the potential to reduce the lifetimes of wall components and even to destroy sensitive, i.e. actively cooled parts. The research for effective ways to suppress the generation of runaway electrons is hampered by the lack of a complete understanding of the physics of the runaways in disruptions. As it is practically impossible to use standard electron detectors in the challenging environment of a tokamak, the experimental knowledge about runaways is limited and it relies on rather indirect techniques of measurement. The main diagnostics used for this PhD work are three reciprocating probes which measure the runaway electrons directly at the plasma edge of the tokamak TEXTOR. A calorimetric probe and a material probe which exploits the signature that a runaway beam impact leaves in the probe were developed in the course of the PhD work. Novel observations of the burst-like runaway electron losses in tokamak disruptions are reported. The runaway bursts are temporally resolved and first-time measurements of the corresponding runaway energy spectra are presented. A characteristic shape and typical burst to burst variations of the

  7. Plasmator. A numerical code for simulation of plasma transport in Tokamaks

    International Nuclear Information System (INIS)

    Guasp, J.

    1979-01-01

    Plasmator is a flexible monodimensional numerical code for plasma transport in Tokamaks of circular cross-section, it allows neutral particle transport and impurity effects. The code leaves a total freedom in the analytical form of transport coefficients. It has been writen in Fortran-V for the UNIVAC-1100/80 from JEN and allows for the possibility of graphics for radial profiles and temporal evolution of the main plasma magnitudes, as well in three-dimensional as in two-dimensional representation either on a Calcomp plotter or in the printer. (author)

  8. Importance of Plasma Response to Non-axisymmetric Perturbations in Tokamaks

    International Nuclear Information System (INIS)

    Park, Jong-kyu; Boozer, Allen H.; Menard, Jonathan E.; Garofalo, Andrea M.; Schaffer, Michael J.; Hawryluk, Richard J.; Kaye, Stanley M.; Gerhardt, Stefan P.; Sabbagh, Steve A. and the NSTX Team

    2009-01-01

    Tokamaks are sensitive to deviations from axisymmetry as small as (delta)B/B 0 ∼ 10 -4 . These non-axisymmetric perturbations greatly modify plasma confinement and performance by either destroying magnetic surfaces with subsequent locking or deforming magnetic surfaces with associated non-ambipolar transport. The Ideal Perturbed Equilibrium Code (IPEC) calculates ideal perturbed equilibria and provides important basis for understanding the sensitivity of tokamak plasmas to perturbations. IPEC calculations indicate that the ideal plasma response, or equivalently the effect by ideally perturbed plasma currents, is essential to explain locking experiments on National Spherical Torus eXperiment (NSTX) and DIII-D. The ideal plasma response is also important for Neoclassical Toroidal Viscosity (NTV) in non-ambipolar transport. The consistency between NTV theory and magnetic braking experiments on NSTX and DIII-D can be improved when the variation in the field strength in IPEC is coupled with generalized NTV theory. These plasma response effects will be compared with the previous vacuum superpositions to illustrate the importance. However, plasma response based on ideal perturbed equilibria is still not sufficiently accurate to predict the details of NTV transport, and can be inconsistent when currents associated with a toroidal torque become comparable to ideal perturbed currents

  9. Problems with the concept of plasma equilibrium in tokamaks

    International Nuclear Information System (INIS)

    Carreras, B.A.

    1992-01-01

    The equilibrium condition for a magnetically confined plasma in normally formulated in terms of macroscopic equations. In these equations, the plasma pressure is assumed to be a function of the magnetic flux with continuous derivatives. However, in three- dimensional systems this is not necessarily the case. Here, we look at the case of an intrinsically three-dimensional realistic tokamak, and we discuss the possible interconnection between the equilibrium and anomalous transport

  10. Thermonuclear Tokamak plasmas in the presence of fusion alpha particles

    International Nuclear Information System (INIS)

    Anderson, D.; Hamnen, H.; Lisak, M.

    1988-01-01

    In this overview, we have focused on several results of the thermonuclear plasma research pertaining to the alpha particle physics and diagnostics in a fusion tokamak plasma. As regards the discussion of alpha particle effects, two distinct classes of phenomena have been distinguished: the simpler class containing phenomena exhibited by individual alpha particles under the influence of bulk plasma properties and, the more complex class including collective effects which become important for increasing alpha particle density. We have also discussed several possibilities to investigate alpha particle effects by simulation experiments using an equivalent population of highly energetic ions in the plasma. Generally, we find that the present theoretical knowledge on the role of fusion alpha particles in a fusion tokamak plasma is incomplete. There are still uncertainties and partial lack of quantitative results in this area. Consequently, further theoretical work and, as far a possible, simulation experiments are needed to improve the situation. Concerning the alpha particle diagnostics, the various diagnostic techniques and the status of their development have been discussed in two different contexts: the escaping alpha particles and the confined alpha particles in the fusion plasma. A general conclusion is that many of the different diagnostic methods for alpha particle measurements require further major development. (authors)

  11. Far infrared polarimetry with tokamak plasmas for determination of the poloidal magnetic field distribution

    Energy Technology Data Exchange (ETDEWEB)

    Kunz, W

    1979-01-01

    This study examines the poloidal magnetic field distribution of tokamak plasma, and the elucidation of the radial distribution of the toroidal plasma flow. A numerical and experimental determination of the poloidal field based on the Faraday effect is presented. A method is discussed for measuring the rotation of the polarization plane linear polarized electromagnetic radiation, by passing through a plasma magnetized in the direction of the radiation. The polarization behavior of a linear polarized wave passing through a tokamak plasma is presented theoretically for various wavelengths, along with the experimental investigation of a ferrite modulation procedure through the use of different far infrared detectors.

  12. 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 obtain...... for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath....

  13. Experimental observation of current generation by asymmetrical heating of ions in a tokamak plasma

    International Nuclear Information System (INIS)

    Gahl, J.; Ishihara, O.; Wong, K.L.; Kristiansen, M.; Hagler, M.

    1986-01-01

    The first experimental observation of current generation by asymmetrical heating of ions is reported. Ions were asymmetrically heated by a unidirectional fast Alfven wave launched by a slow wave antenna inside a tokamak. Current generation was detected by measuring the asymmetry of the toroidal plasma current with probes at the top and bottom of the toroidal plasma column

  14. Study of the heavy impurity influx into the plasma using laser fluorescence spectroscopy in the TO-2 tokamak with toroidal divertor

    International Nuclear Information System (INIS)

    Vukolov, K.Yu.; Shvindt, N.N.

    1992-01-01

    Measurement cycle for determination of iron atom absolute concentrations was carried out in divertor and diaphragm modes of laser fluorescence spectroscopy. The conclusion is made on effective wall shielding by divertor layer as compared to material diaphragm. The basic result of the work consists in creating and testing on the tokamak TO-2 of multichannel diagnostic complex for remote measurement of atom (ion) absolute concentrations of metallic impurities in the near-wall plasma with high spatial and time resolution through laser fluorescence spectroscopy method intended for studies at the Tokamak-15 facility

  15. Injection of electrons with predominantly perpendicular energy into an area of toroidal field ripple in a tokamak plasma to improve plasma confinement

    Science.gov (United States)

    Ono, Masayuki; Furth, Harold

    1993-01-01

    An electron injection scheme for controlling transport in a tokamak plasma. Electrons with predominantly perpendicular energy are injected into a ripple field region created by a group of localized poloidal field bending magnets. The trapped electrons then grad-B drift vertically toward the plasma interior until they are detrapped, charging the plasma negative. Calculations indicate that the highly perpendicular velocity electrons can remain stable against kinetic instabilities in the regime of interest for tokamak experiments. The penetration distance can be controlled by controlling the "ripple mirror ratio", the energy of the injected electrons, and their v.sub..perp. /v.sub.51 ratio. In this scheme, the poloidal torque due to the injected radial current is taken by the magnets and not by the plasma. Injection is accomplished by the flat cathode containing an ECH cavity to pump electrons to high v.sub..perp..

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

    International Nuclear Information System (INIS)

    Nam, Y. U.; Chung, J.

    2010-01-01

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

  17. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    Amiel, Stephane

    2014-01-01

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

  18. Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Bell, M.G.; Beer, M.

    1997-02-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l i ). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q a ∼ 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l i plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q 0 > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions

  19. Observation of E×B Flow Velocity Profile Change Using Doppler Reflectometry in HL-2A

    Institute of Scientific and Technical Information of China (English)

    XIAO Wei-Wen; ZOU Xiao-Lan; DING Xuan-Tong; DONG Jia-Qi; LIU Ze-Tian; SONG Shao-Dong; GAO Ya-Dong; YAO Liang-Hua; FENG Bei-Bin; SONG Xian-Ming; CHEN Cheng-Yuan; SUN Hong-Juan; LI Yong-Gao; YANG Qing-Wei; YAN Long-Wen; LIU Yi; DUAN Xu-Ru; PAN Chuan-Hong; LIU Yong

    2009-01-01

    A broadband,O-mode sweeping Doppler reflectometry designed for measuring plasma E×B flow velocity profiles is operated in HL-2A.The main feature of the Doppler reflectometry is its capability to be tuned to any selected frequency in total waveband from 26-40 GHz.This property enables us to probe several plasma layers within a short time interval during a discharge,permitting the characterization of the radial distribution of plasma fluctuations.The system allows us to extract important information about the velocity change layer,namely its spatial localization.In purely Ohmic discharge a change of the E×B flow velocity profiles has been observed in the region for 28 < r < 30cm if only the line average density exceeds 2.2×1019 m-3.The density gradient change is measured in the same region,too.

  20. Metal droplet erosion and shielding plasma layer under plasma flows typical of transient processes in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Martynenko, Yu. V., E-mail: Martynenko-YV@nrcki.ru [National Research Nuclear University “MEPhI” (Russian Federation)

    2017-03-15

    It is shown that the shielding plasma layer and metal droplet erosion in tokamaks are closely interrelated, because shielding plasma forms from the evaporated metal droplets, while droplet erosion is caused by the shielding plasma flow over the melted metal surface. Analysis of experimental data and theoretical models of these processes is presented.

  1. Rapid further heating of tokamak plasma by fast-rising magnetic pulse

    International Nuclear Information System (INIS)

    Inoue, N.; Nihei, H.; Yamazaki, K.; Ichimura, M.; Morikawa, J.; Hoshino, K.; Uchida, T.

    1977-01-01

    The object of the experiment was to study the rapid further heating of a tokamak plasma and its influence on confinement. For this purpose, a high-voltage theta-pinch pulse was applied to a tokamak plasma and production of a high-temperature (keV) plasma was ensured within a microsecond. The magnetic pulse is applied at the plasma current maximum parallel or antiparallel to the study toroidal field. In either case, the pulsed field quickly penetrates the plasma and the plasma resistivity estimated from the penetration time is about 100 times larger than the classical. A burst of energetic neutrals of approximately 1 μs duration was observed and the energy distribution had two components of the order of 1 keV and 0.1 keV in the antiparallel case. Doppler broadening measurement shows heating of ions to a temperature higher than 200 eV; however, the line profile is not always Maxwellian distribution. The X-rays disappear at the moment of applying the magnetic pulse and reappear about 100 μs later with an intensive burst, while both energy levels are the same (approximately 100 keV). (author)

  2. Preliminary experiment of non-induced plasma current startup on SUNIST spherical tokamak

    International Nuclear Information System (INIS)

    He Yexi; Zhang Liang; Xie Lifeng; Tang Yi; Yang Xuanzong; Fu Hongjun

    2005-01-01

    Non-inductive plasma current startup is an important motivation on the SUNIST spherical tokamak. In this experiment, a 100 kW, 2.45 GHz magnetron microwave system has been applied to the plasma current startup. Besides the toroidal field, a vertical field was applied to generate a preliminary toroidal plasma current without action of the central solenoid. As the evidence of the plasma current startup by the vertical field drift effect, the direction of the plasma current is changed with the changing direction of the vertical field during ECR startup discharge. We have also observed the plasma current maximum by scanning the vertical field in both directions. Additionally, we have used electrode discharge to assist the ECR current startup. (author)

  3. Static and dynamic control of plasma equilibrium in a Tokamak

    International Nuclear Information System (INIS)

    Blum, J.; Dei Cas, R.

    1979-01-01

    We are dealing here with the problem of controlling the plasma boundary and its displacements. Static control consists in determining the currents in the external coils of the Tokamak so that the plasma boundary has certain fixed characteristics: radial position, vertical elongation, desired shape. A self-consistent method is proposed here, considering a free plasma boundary, and using the techniques of optimal control of distributed parameter systems to solve the problem. The dynamic control problem considered in the second part of the paper is the control of the plasma radial displacements. An elaborate system of preprogramming and feedback control has been developed to ensure equilibrium and stability of the horizontal plasma motions. Optimal control techniques have been used to calculate the optimal primary coils configuration, the preprogramming voltages and the feedback gains. A new stability diagrams has been obtained which takes into account the erosion of the plasma by the limiter. All these calculations have been applied successfully to TFR 600 where thin liner and the presence of an iron core make the problem of stabilization of the radial displacements very difficult

  4. DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS

    International Nuclear Information System (INIS)

    LUCE, T.C.; WADE, M.R.; FERRON, J.R.; HYATT, A.W.; KELLMAN, A.G.; KINSEY, J.E.; LAHAY, R.J.; LASNIER, C.J.; MURAKAMI, M.; POLITZER, P.A.; SCOVILLE, J.T.

    2002-01-01

    OAK A271 DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS. Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak in stationary conditions with relatively low plasma current (q 95 > 4). A figure of merit for fusion gain Β N H 89 /q 95 2 has been maintained at values corresponding to Q = 10 operation in a burning plasma for > 6 s or 36 τ E and 2 τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1, which allows stable operation up to the no-wall ideal β limit. These plasmas maintain particle balance by active pumping rather than transient wall conditions. The reduced current lessens significantly the potential for structural damage in the event of a major disruption

  5. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  6. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  7. Transport Bifurcation in a Rotating Tokamak Plasma

    International Nuclear Information System (INIS)

    Highcock, E. G.; Barnes, M.; 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 nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

  8. Electron temperature gradient driven instability in the tokamak boundary plasma

    International Nuclear Information System (INIS)

    Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.

    1992-01-01

    A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t -1/2 e γmt

  9. Plasma fluctuation measurements in tokamaks using beam-plasma interactions (abstract)

    International Nuclear Information System (INIS)

    Fonck, R.J.; Duperrex, P.A.; Paul, S.F.

    1990-01-01

    High-frequency observations of light emitted from the interactions between plasma ions and injected neutral beam atoms allow the measurement of moderate-wavelength fluctuations in plasma and impurity ion densities. To detect turbulence in the local plasma ion density, the collisionally excited fluorescence from a neutral beam is measured either separately at several spatial points or with a multichannel imaging detector. Similarly, the role of impurity ion density fluctuations is measured using charge exchange recombination excited transitions emitted by the ion species of interest. This technique can access the relatively unexplored region of long-wavelength plasma turbulence with k perpendicular ρ i much-lt 1, and hence complements measurements from scattering experiments. Optimization of neutral beam geometry and optical sightlines can result in very good localization and resolution (Δx≤1 cm) in the hot plasma core region. The detectable fluctuation level is determined by photon statistics, atomic excitation processes, and beam stability, but can be as low as 0.2% in a 100 kHz bandwidth over the 0--1 MHz frequency range. The choices of beam species (e.g., H 0 , He 0 , etc.), observed transition (e.g., H α , L α , He I singlet or triplet transitions, C VI Δn=1, etc.) are dictated by experiment-specific factors such as optical access, flexibility of beam operation, plasma conditions, and detailed experimental goals. Initial tests on the PBX-M tokamak using the H α emissions from a heating neutral beam show low-frequency turbulence in the edge plasma region

  10. ISTTOK tokamak plasmas influence on a liquid gallium jet dynamic behavior

    International Nuclear Information System (INIS)

    Gomes, R.B.; Silva, C.; Fernandes, H.; Duarte, P.; Nedzelskiy, I.; Lielausis, O.; Klyukin, A.; Platacis, E.

    2011-01-01

    The main concern in using free flowing liquid metals in fusion devices is related to their interaction with magnetic fields. On ISTTOK tokamak, liquid gallium jets are injected deep into the plasma along a vertical direction. The influence of the plasma interaction on the jet has been investigated monitoring the liquid metal behavior using a fast frame camera. A radial shift on its trajectory has been detected and found to depend on the toroidal magnetic field magnitude and principally on the plasma position within the chamber. The analysis performed to understand the dynamics of the jet perturbation by the plasma is presented in this paper. The jet surface temperature increase during this interaction has also been measured, using absolutely calibrated multichannel IR sensors, to evaluate the jet power exhaustion capability.

  11. ISTTOK tokamak plasmas influence on a liquid gallium jet dynamic behavior

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, R.B., E-mail: gomes@ipfn.ist.utl.pt [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, 1049-001 Lisboa (Portugal); Silva, C.; Fernandes, H.; Duarte, P.; Nedzelskiy, I. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, 1049-001 Lisboa (Portugal); Lielausis, O.; Klyukin, A.; Platacis, E. [Association EURATOM/University of Latvia, Institute of Solid State Physics, 8 Kengaraga Str., LV-1063 Riga (Latvia)

    2011-08-01

    The main concern in using free flowing liquid metals in fusion devices is related to their interaction with magnetic fields. On ISTTOK tokamak, liquid gallium jets are injected deep into the plasma along a vertical direction. The influence of the plasma interaction on the jet has been investigated monitoring the liquid metal behavior using a fast frame camera. A radial shift on its trajectory has been detected and found to depend on the toroidal magnetic field magnitude and principally on the plasma position within the chamber. The analysis performed to understand the dynamics of the jet perturbation by the plasma is presented in this paper. The jet surface temperature increase during this interaction has also been measured, using absolutely calibrated multichannel IR sensors, to evaluate the jet power exhaustion capability.

  12. Feedback stabilization of the axisymmetric instability of a deformable tokamak plasma

    International Nuclear Information System (INIS)

    Pomphrey, N.; Jardin, S.C.; Ward, D.J.

    1989-01-01

    The paper presents an analysis of the magnetohydrodynamic stability of the axisymmetric system consisting of a free boundary tokamak plasma with non-circular cross-section, finite resistivity passive conductors, and an active feedback system with magnetic flux pickup loops, a proportional amplifier with gain G and current carrying poloidal field coils. A numerical simulation of the system when G is set to zero identifies flux loop locations which correctly sense the plasma motion. However, when certain of these locations are incorporated into an active feedback scheme, the plasma fails to be stabilized, no matter what value of the gain is chosen. Analysis on the basis of an extended energy principle indicates that this failure is due to the deformability of the plasma cross-section. (author). 14 refs, 7 figs

  13. Development of fast video recording of plasma interaction with a lithium limiter on T-11M tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, V.B., E-mail: v_lazarev@triniti.ru [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Dzhurik, A.S.; Shcherbak, A.N. [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Belov, A.M. [NRC “Kurchatov Institute”, Moscow (Russian Federation)

    2016-11-15

    Highlights: • The paper presents the results of the study of tokamak plasma interaction with lithium capillary-porous system limiters and PFC by high-speed color camera. • Registration of emission near the target in SOL in neutral lithium light and e-folding length for neutral Lithium measurements. • Registration of effect of MHD instabilities on CPS Lithium limiter. • A sequence of frames shows evolution of lithium bubble on the surface of lithium limiter. • View of filament structure near the plasma edge in ohmic mode. - Abstract: A new high-speed color camera with interference filters was installed for fast video recording of plasma-surface interaction with a Lithium limiter on the base of capillary-porous system (CPS) in T-11M tokamak vessel. The paper presents the results of the study of tokamak plasma interaction (frame exposure time up to 4 μs) with CPS Lithium limiter in a stable stationary phase, unstable regimes with internal disruption and results of processing of the image of the light emission around the probe, i.e. e-folding length for neutral Lithium penetration and e-folding length for Lithium ion flux in SOL region.

  14. Experiment and research on materials irradiated by plasma radiation

    International Nuclear Information System (INIS)

    Hong Wenyu; Yao Lianghua; Tang Sujun; Chang Shufen; Li Guodong

    1992-08-01

    The TiC and SiC coating on the graphite substrate and wall carbonization were studied by plasma radiation in HL-1 tokamak. Samples were analysed with AES (auger electron spectroscopy), SEM (scanning electron microscopy), XPS (X-ray photoelectron spectroscopy) and XDS (X-ray diffraction spectroscopy). The results show that the TiC and SiC materials coated on limiter and wall and wall carbonization can reduce the metal and oxygen impurities and improve the plasma merit

  15. Edge radial electric field structure in quiescent H-mode plasmas in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); West, W P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Doyle, E J [University of California, Los Angeles, CA 90095-1597 (United States); Austin, M E [University of Texas at Austin, Austin, TX 78712 (United States); DeGrassie, J S [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Gohil, P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Greenfield, C M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Jayakumar, R [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); Kaplan, D H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lao, L L [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M A [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); McKee, G R [University of Wisconsin, Madison, WI 53706-1687 (United States); Solomon, W M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Rhodes, T L [University of California, Los Angeles, CA 90095-1597 (United States); Wade, M R [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wang, G [University of California, Los Angeles, CA 90095-1597 (United States); Watkins, J G [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Zeng, L [University of California, Los Angeles, CA 90095-1597 (United States)

    2004-05-01

    H-mode operation is the choice for next step tokamak devices based on either conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the {beta} limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D over the past four years have demonstrated a new operating regime, the quiescent H-mode (QH-mode) regime, that solves these problems. QH-mode plasmas have now been run for over 4 s (>30 energy confinement times). Utilizing the steady-state nature of the QH-mode edge allows us to obtain unprecedented spatial resolution of the edge ion profiles and the edge radial electric field, E{sub r}, by sweeping the edge plasma slowly past the view points of the charge exchange spectroscopy system. We have investigated the effects of direct edge ion orbit loss on the creation and sustainment of the QH-mode. Direct loss of ions injected into the velocity-space loss cone at the plasma edge is not necessary for creation or sustainment of the QH-mode. The direct ion orbit loss has little effect on the edge E{sub r} well. The E{sub r} at the bottom of the well in these cases is about -100 kV m{sup -1} compared with -20 to -30 kV m{sup -1} in the standard H-mode. The well is about 1 cm wide, which is close to the diameter of the deuteron gyro-orbit. We also have investigated the effect of changing edge triangularity by changing the plasma shape from upwardly biased single null to magnetically balanced double null. We have now achieved the QH-mode in these double-null plasmas. The increased triangularity allows us to increase pedestal density in QH-mode plasmas by a factor of about 2.5 and overall pedestal pressure by a factor of 2. Pedestal {beta} and {nu}{sup *} values matching the values desired for ITER have been achieved. In

  16. On lateral deflection of the SOL plasma in tokamaks during giant ELMs

    International Nuclear Information System (INIS)

    Landman, I.S.; Wuerz, H.

    2000-06-01

    In recent H-mode experiments at JET with giant ELMs a lateral deflection of hot tokamak plasma leaving the scrape-off layer and striking the divertor plate has been observed. This deflection can effect the divertor erosion caused by the hot plasma irradiation, because of enlarging the irradiated area. A simplified MHD model of the vapor shield plasma and of the hot plasma initially formed at time t → -∞ is analyzed. At t = -∞ both plasmas are assumed to stay on rest and to be separated by a boundary, which is parallel to the plate surface. The interaction between plasmas is assumed to develop gradually ('adiabatically') as exp(t/t 0 ) with t 0 ∝ 10 2 μs the ELM duration time. Electrical insulation of the core tokamak plasma is assumed everywhere except for the contact with the divertor. Electric currents are flowing only in the toroidal direction. These currents developing in the interaction zone of the hot plasma and the rather cold target plasma are calculated for inclined impact of the magnetized hot plasma. At such conditions the J x B force in the lateral direction accelerates the interacting plasmas. The motion of the cold plasma and the gradual increase of the plasma interaction intensity are shown to be important for the appropriate deflection magnitude. Adiabatically responding against the increase of the interaction intensity the cold plasma motion compensates significantly the currents thus decreasing the deflection compared to motionless approach. The calculated magnitude of the hot plasma deflection is comparable to the observed one. The results of the modeling are discussed in relation to the experiments. It is shown that sudden switching on of the interaction produces Alfven oscillations of large amplitudes causing much larger amplitudes of the magnetic field induced by the currents than in the adiabatic case. (orig.)

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

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio

    1997-01-01

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

  18. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-06-01

    The objective of this thesis is the study of plasma microinstabilities in linear and nonlinear tokamak regime. After a brief review of experimental results the theoretical tools used in this study are presented. A variational method founded on the existence of angular variables system and on action for charged particles in tokamak configurations is detailed. The correspondent functional extreme with regard to fluctuating electromagnetic field, is calculated analytically with taking into account the toroidal geometry. A numerical code, TORRID, has been constructed on this principle and the main instabilities, particularly ionic instabilities and microtearing, has been linearly studied. The most simple non linear methods are rewieved and applied at the microtearing instabilities. The quasilinear transport coefficients are deducted of an entropy minimum production principle. The ionic thermic conductivity and the viscosity are calculated for an ionic turbulence [fr

  19. Thermal loads on tokamak plasma-facing components during normal operation and disruptions

    International Nuclear Information System (INIS)

    McGrath, R.T.

    1990-01-01

    Power loadings experienced by tokamak plasma-facing components during normal operation and during off-normal events are discussed. A model for power and particle flow in the tokamak boundary layer is presented and model predictions are compared to infrared measurements of component heating. The inclusion of the full three-dimensional geometry of the components and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armour tile surfaces by only a millimeter can lead to significant localized heating. An application to the design of plasma-facing components for future machines is presented. Finally, thermal loads expected during tokamak disruptions are discussed. The primary problems are surface melting and vaporization due to localized intense heating during the disruption thermal quench and volumetric heating of the component armour and structure due to localised impact of runaway electrons. (author)

  20. On-line system for control of plasma filament position in the Tokamak-10

    International Nuclear Information System (INIS)

    Britousov, N.N.; Valuev, S.F.; Sychev, G.I.; Shchedrov, V.M.

    1982-01-01

    The plasma filament position on-line control system (OCS) in the T-10 tokamak is described. Results of adjustment and operation of the system are given. The OCS is a structure of a direct negative feedback (DNF) versus deflection and a local DNF circuit. The OCS experimental studying is carried out under the following conditions: 200 kA plasma current, 32 cm diaphragm radius, 2.2-2.5 stability margin, 440 V anode voltage. The response time for 2 cm deflection jumps is 15-20 ns. The OCS demonstrated a particular efficiency while operating in parallel with the plasma current stabilizer providing a high discharge repetition and considerably reducing the number of substandard pulses

  1. Comments on experimental results of energy confinement of tokamak plasmas

    International Nuclear Information System (INIS)

    Chu, T.K.

    1989-04-01

    The results of energy-confinement experiments on steady-state tokamak plasmas are examined. For plasmas with auxiliary heating, an analysis based on the heat diffusion equation is used to define heat confinement time (the incremental energy confinement time). For ohmically sustained plasmas, experiments show that the onset of the saturation regime of energy confinement, marfeing, detachment, and disruption are marked by distinct values of the parameter /bar n//sub e///bar j/. The confinement results of the two types of experiments can be described by a single surface in 3-dimensional space spanned by the plasma energy, the heating power, and the plasma density: the incremental energy confinement time /tau//sub inc/ = ΔW/ΔP is the correct concept for describing results of heat confinement in a heating experiment; the commonly used energy confinement time defined by /tau//sub E/ = W/P is not. A further examination shows that the change of edge parameters, as characterized by the change of the effective collision frequency ν/sub e/*, governs the change of confinement properties. The totality of the results of tokamak experiments on energy confinement appears to support a hypothesis that energy transport is determined by the preservation of the pressure gradient scale length. 70 refs., 6 figs., 1 tab

  2. Information content of transient synchrotron radiation in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1989-04-01

    A brief, deliberate, perturbation of hot tokamak electrons produces a transient, synchrotron radiation signal, in frequency-time space, with impressive informative potential on plasma parameters; for example, the dc toroidal electric field, not available by other means, may be measurably. Very fast algorithms have been developed, making tractable a statistical analysis that compares essentially all parameter sets that might possibly explain the transient signal. By simulating data numerically, we can estimate the informative worth of data prior to obtaining it. 20 refs., 2 figs

  3. Calculations of axisymmetric stability of tokamak plasmas with active and passive feedback

    International Nuclear Information System (INIS)

    Ward, D.J.; Jardin, S.C.; Cheng, C.Z.

    1991-07-01

    A new linear MHD stability code, NOVA-W, has been developed in order to study feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The passive stability predictions of the code have been tested both against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops that provide the measurements of vertical displacement. The results compare well with those computed earlier using a less efficient nonlinear code. 37 refs., 13 figs

  4. Numerical study of the electron heating and current drive by the fast waves in the JFT-2M tokamak plasma

    International Nuclear Information System (INIS)

    Yamamoto, Takumi; Uesugi, Yoshihiko; Hoshino, Katsumichi; Kawashima, Hisato; Ohtsuka, Hideo

    1986-08-01

    A 200 MHz fast wave experiment for the JET-2M tokamak is examined. Noticeable single-path electron Landau damping of the fast waves with the parallel refractive index of N // = 4 is expected in the plasma with electron temperature more than 2.5 keV at the electron density of n e = 1.5 x 10 19 m -3 . Furthermore, it is shown that 8 kA of the plasma current is driven by the fast waves with N //≅ 2 at n e = 3 x 10 19 m -3 in the single-path damping when 100 kW of the rf power radiates into the plasma in the presence of the hot electrons with the temperature of 19 keV and the fraction of the density of 2 %. (author)

  5. Magnetohydrodynamic equilibria and local stability of axisymmetric tokamak plasmas

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Dory, R.A.; Nelson, D.B.; Sayer, R.O.

    1976-07-01

    Axisymmetric magnetohydrodynamic equilibria are evaluated in terms of the Mercier Stability Criterion. The parameters of interest include poloidal beta (β/sub p/), current and pressure profile widths, D-shaped and doublet plasmas with elongation (sigma) and triangularity (delta), and the aspect ratio (A). For marginal local stability, the critical values of β, plasma current, and the safety factor q with fixed toroidal field at the geometric center of the plasma are obtained. It is shown that for a wide range of profiles in a D-shaped plasma with A = 3, the highest critical β occurs at β/sub p/ = 2.4, sigma = 1.65, and delta = 0.5. If the toroidal field at the coil surface is fixed, the highest critical pressure occurs near A approximately 3 to 4, given reasonable distance between the coils and the plasma edge. Calculations for a Doublet II-A plasma with sigma = 3 show that with similar pressure profile the highest critical β occurs at β/sub p/ = 1 and is 84 percent of the highest critical β for the D-shaped plasmas. Critical values of ohmic heating power density are also found to be comparable for the two plasma shapes. A D-shaped plasma with the above parameters is suggested for use in future high-β tokamak devices

  6. 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.

  7. A dynamic state observer for real-time reconstruction of the tokamak plasma profile state and disturbances

    NARCIS (Netherlands)

    Felici, F.; De Baar, M.; Steinbuch, M.

    2014-01-01

    A dynamic observer is presented which can reconstruct the internal state of a tokamak fusion plasma, consisting of the spatial distribution of current and temperature, from measurements. Today, the internal plasma state is usually reconstructed by solving an ill-conditioned inversion problem using a

  8. Development of a six channel Fabry-Perot interferometer for continuous measurement of electron temperature of Tokamak plasma. Application to current diffusion study

    International Nuclear Information System (INIS)

    Talvard, M.

    1984-10-01

    It is shown how the properties of the electron cyclotron emission of a tokamak plasma can be used to measure the electron temperature. The design of a six channel Fabry-Perot interferometer is then described. This interferometer allows the measurement of the time evolution of the electron temperature profile of the plasma in the TFR tokamak. Using this technique interesting results have been obtained concerning the current penetration during the start up phase of a tokamak discharge [fr

  9. Simulation of MHD instability effects on burning plasma transport with ITB in tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Yamada, I.; Taniguchi, S.; Oishi, T.

    2009-01-01

    Full text: The high performance plasma behavior is required to realize economic and environmental-friendly fusion reactors compatible with conventional power plant systems. To improve plasma confinement, the formation of internal transport barrier (ITB) is anticipated, and its behavior is analyzed by the simulation code TOTAL (Toroidal Transport Linkage Analysis). This TOTAL code comprises a 2- or 3-dimensional equilibrium and 1-dimensional predictive transport code for both tokamak and helical systems. In the tokamak code TOTAL-T, the external current drive, bootstrap current, sawtooth oscillation, ballooning mode and neoclassical tearing mode (NTM) analyses are included. The steady-state burning plasma operation is achieved by the feedback control of pellet injection fuelling and external heating power control. The impurity dynamics of iron and tungsten is also included in this code. The NTM effects are evaluated using the modified Rutherford Model with the stabilization of the ECCD current drive. The excitation of m=2/n=1 NTM leads to the 20 % reduction in the central temperature in ITER-like reactors. Recently, the external non-resonant helical field application is analyzed and its stabilization properties are evaluated. The pellet injection effects on ITB formation is also clarified in tokamak and helical plasmas. Relationship between sawtooth oscillation and impurity ejection is recently simulated in comparison with experimental data. In this conference, we will show above-stated new results on MHD instability effects on burning plasma transport. (author)

  10. Fokker--Planck/transport analyses of fusion plasmas in contemporary beam-driven tokamaks

    International Nuclear Information System (INIS)

    Mirin, A.A.; McCoy, M.G.; Killeen, J.; Rensink, M.E.; Shumaker, D.E.; Jassby, D.L.; Post, D.E.

    1978-04-01

    The properties of deuterium plasmas in experimental tokamaks heated and fueled by intense neutral-beam injection are evaluated with a Fokker-Planck/radial transport code coupled with a Monte Carlo neutrals treatment. Illustrative results are presented for the Poloidal Divertor Experiment at PPPL as a function of beam power and plasma recycling coefficient, R/sub c/. When P/sub beam/ = 8 MW at E/sub b/ = 60 keV, and R/sub c/ = 0.2, then approximately 0.5, [ 2 / 3 ] = 22 keV approximately 6 , and the D-D neutron intensity is 10 16 n/sec

  11. Theory-based scaling of the SOL width in circular limited tokamak plasmas

    International Nuclear Information System (INIS)

    Halpern, F.D.; Ricci, P.; Labit, B.; Furno, I.; Jolliet, S.; Loizu, J.; Mosetto, A.; Arnoux, G.; Silva, C.; Gunn, J.P.; Horacek, J.; Kočan, M.; LaBombard, B.

    2013-01-01

    A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement. (letter)

  12. Calculation of poloidal rotation in the edge plasma of limiter tokamaks

    International Nuclear Information System (INIS)

    Gerhauser, H.; Claassen, H.A.

    1987-05-01

    The existing 2-d two-fluid code for computing the plasma profiles in the scrape-off layer of limiter tokamaks has been further developed to include the effect of poloidal rotation in the basic equations. This rotation is produced by radial electric fields which arise in the limiter shadow due to radial gradients in the Langmuir sheath potential in front of the limiter. As a consequence slight deviations from ambipolar motion must occur. A strong increase of rotation near the separatrix is connected with an electric current circuit closed via the limiter edge. The 2-d profiles of all relevant quantities are calculated and discussed for TEXTOR-typical parameters including also the effect of limiter recycled neutrals. The results agree well with the known experimental evidence on poloidal rotation and should be transferable to all limiter tokamaks. (orig.)

  13. Comparison of wall/divertor deuterium retention and plasma fueling requirements on the DIII-D, TdeV, and ASDEX-upgrade tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge Associated Universities, TN (United States); Terreault, B. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada); Haas, G. [Max Planck Inst. fuer Plasmaphysik, Garching (Germany)] [and others

    1996-06-01

    The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.

  14. Electromagneto-mechanical coupling analysis of a test module in J-TEXT Tokamak during plasma disruption

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Haijie; Yuan, Zhensheng; Yuan, Hongwei; Pei, Cuixiang [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Chen, Zhenmao, E-mail: chenzm@mail.xjtu.edu.cn [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Yang, Jinhong; Wang, Weihua [Institute of Applied Physics of AOA, Hefei 230031 (China)

    2016-11-01

    In this paper, the dynamic response during plasma disruption of a test blanket module in vacuum vessel (VV) of the Joint TEXT (J-TEXT), which is an experimental Tokamak device with iron core, was simulated by applying a program developed by authors on the ANSYS platform using its parametric design language (APDL). The moving coordinate method as well as the load transfer and sequential coupling strategy were adopted to cope with the electromagneto-mechanical coupling effect. To establish the numerical model, the influence of the iron core on the eddy current and electromagnetic (EM) force during disruption was numerically investigated at first and the influence was found not significant. Together with the geometrical features of the J-TEXT Tokamak structure, 180° sector models without magnetic core were finally established for the EM field and the structural response simulations. To obtain the source plasma current, the plasma current evolution during disruption was simulated by using the Tokamak Simulation Code (TSC). With the numerical models and the source plasma current, the dynamic response of both the VV structure and the test module were calculated. The numerical results show that the maximum stress of the test module is in safe range, and the magnetic damping effect can weaken vibration of the test module. In addition, simulation without considering the coupling effect was carried out, which shows that the influence of coupling effect is not significant for the peak stress of the J-TEXT disruption problem.

  15. Impurity flux collection at the plasma edge of the tokamak MT-1

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Bakos, J.S.; Petravich, G.

    1989-09-01

    Fluxes of intrinsic and injected impurities and background plasma ions were collected using a bidirectional probe at the plasma edge of the tokamak MT-1. The directional and radial dependences of injected impurities and plasma ions were very similar indicating a strong coupling of the impurity transport to the dynamics of the background plasma. The measured intrinsic concentration of about 10 -4 for Mo at the plasma edge is derived. (author) 17 refs.; 5 figs

  16. MAIA, Eigenvalues for MHD Equation of Tokamak Plasma Stability Problems

    International Nuclear Information System (INIS)

    Tanaka, Y.; Azumi, M.; Kurita, G.; Tsunematsu, T.; Takeda, T.

    1986-01-01

    1 - Description of program or function: This program solves an eigenvalue problem zBx=Ax where A and B are real block tri-diagonal matrices. This eigenvalue problem is derived from a reduced set of linear resistive MHD equations which is often employed to study tokamak plasma stability problem. 2 - Method of solution: Both the determinant and inverse iteration methods are employed. 3 - Restrictions on the complexity of the problem: The eigenvalue z must be real

  17. Kinetic theory of plasma adiabatic major radius compression in tokamaks

    International Nuclear Information System (INIS)

    Gorelenkova, M.V.; Gorelenkov, N.N.; Azizov, E.A.; Romannikov, A.N.; Herrmann, H.W.

    1998-01-01

    In order to understand the individual charged particle behavior as well as plasma macroparameters (temperature, density, etc.) during the adiabatic major radius compression (R-compression) in a tokamak, a kinetic approach is used. The perpendicular electric field from the Ohm close-quote s law at zero resistivity is made use of in order to describe particle motion during the R-compression. Expressions for both passing and trapped particle energy and pitch angle change are derived for a plasma with high aspect ratio and circular magnetic surfaces. The particle behavior near the passing trapped boundary during the compression is studied to simulate the compression-induced collisional losses of alpha particles. Qualitative agreement is obtained with the alphas loss measurements in deuterium-tritium (D-T) experiments in the Tokamak Fusion Test Reactor (TFTR) [World Survey of Activities in Controlled Fusion Research [Nucl. Fusion special supplement (1991)] (International Atomic Energy Agency, Vienna, 1991)]. The plasma macroparameters evolution at the R-compression is calculated by solving the gyroaveraged drift kinetic equation. copyright 1998 American Institute of Physics

  18. Role of impurity dynamics in resistivity-gradient-driven turbulence and tokamak edge plasma phenomena

    International Nuclear Information System (INIS)

    Hahm, T.S.; Diamond, P.H.; Terry, P.W.; Garcia, L.; Carreras, B.A.

    1986-03-01

    The role of impurity dynamics in resistivity gradient driven turbulence is investigated in the context of modeling tokamak edge plasma phenomena. The effects of impurity concentration fluctuations and gradients on the linear behavior of rippling instabilities and on the nonlinear evolution and saturation of resistivity gradient driven turbulence are studied both analytically and computationally. At saturation, fluctuation levels and particle and thermal diffusivities are calculated. In particular, the mean-square turbulent radial velocity is given by 2 > = (E 0 L/sub s/B/sub z/) 2 (L/sub/eta/ -1 + L/sub z -1 ) 2 . Thus, edged peaked impurity concentrations tend to enhance the turbulence, while axially peaked concentrations tend to quench it. The theoretical predictions are in semi-quantitative agreement with experimental results from the TEXT, Caltech, and Tosca tokamaks. Finally, a theory of the density clamp observed during CO-NBI on the ISX-B tokamak is proposed

  19. Behavior of oxygen impurities in tokamak. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    El-Sharif, R N; Beket, A H [Plasma and Nuclear Fusion Department, Nuclear Research Center, Atomic Energy Aurhority, Cairo (Egypt)

    1996-03-01

    Impurity transport in tokamak plasma is a subject of great importance in present day tokamak experiments. The transport of oxygen as an impurity element in small tokamak was studied theoretically. The viscosity coefficient of oxygen has been calculated in different approximation 13 and 21 moment approximation, taking into consideration {chi}>>1,{chi}{omega}{sub c} {tau}. It was found that in 21 moment approximation additional terms added to the perturbation from equilibrium leads to increase in viscosity coefficients than in 13 moments approximation. 9 figs.

  20. Development of Tokamak experiment technology - Study of ICRF coupling in the KAIST tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Duk In; Chang, Hang Young; Lee, Soon Chil; Kwon, Gi Chung; Seo, Sung Hun; Jeon, Sang Jin; Heo, Sung Hee; Heo, Eun Gi; Lee, Dae Hang; Lee, Chan Hee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-08-01

    Research objectives are to design and fabricate antenna, measure the property of absorption transmitted to the plasma, and research the physical phenomena about the ICRF coupling. Main heating method is ohmic heating at the KAIST tokamak. So, the plasma current produced is more than 30 kA and, the loop voltage of the plasma is 2 {approx} 3V. The power of the plasma by ohmic heating is about 100 kW. Because the toroidal field is 5 {approx} 8 kG, it is needed RF system with more than 100 kW in 7 {approx} 15 MHz. In the first year a RF amplifier with 1 kW in 300 khz {approx} 35 MHz was bought. The manufacture of ICRF system will start from next years. In the research on antenna, we study the method how to measure electric field emitted from antenna using piezo elements. Experimentally, we obtain the results that the signal of piezo element is proportional to the square of electric field. In the next year, we will research the type of antenna subsequently. 28 refs., 3 tabs., 18 figs. (author)

  1. Probiotic assessment of Enterococcus durans 6HL and Lactococcus lactis 2HL isolated from vaginal microflora.

    Science.gov (United States)

    Nami, Yousef; Abdullah, Norhafizah; Haghshenas, Babak; Radiah, Dayang; Rosli, Rozita; Khosroushahi, Ahmad Yari

    2014-08-01

    Forty-five lactic acid bacteria (LAB) were isolated from the vaginal specimens of healthy fertile women, and the identities of the bacteria were confirmed by sequencing of their 16S rDNA genes. Among these bacteria, only four isolates were able to resist and survive in low pH, bile salts and simulated in vitro digestion conditions. Lactococcus lactis 2HL, Enterococcus durans 6HL, Lactobacillus acidophilus 36YL and Lactobacillus plantarum 5BL showed the best resistance to these conditions. These strains were evaluated further to assess their ability to adhere to human intestinal Caco-2 cells. Lactococcus lactis 2HL and E. durans 6HL were the most adherent strains. In vitro tests under neutralized pH proved the antimicrobial activity of both strains. Results revealed that the growth of Escherichia coli O26, Staphylococcus aureus and Shigella flexneri was suppressed by both LAB strains. The antibiotic susceptibility tests showed that these strains were sensitive to all nine antibiotics: vancomycin, tetracycline, ampicillin, penicillin, gentamicin, erythromycin, clindamycin, sulfamethoxazole and chloramphenicol. These data suggest that E. durans 6HL and Lactococcus lactis 2HL could be examined further for their useful properties and could be developed as new probiotics. © 2014 The Authors.

  2. Multi-scale semi-ideal magnetohydrodynamics of a tokamak plasma

    International Nuclear Information System (INIS)

    Bazdenkov, S.; Sato, Tetsuya; Watanabe, Kunihiko.

    1995-09-01

    An analytical model of fast spatial flattening of the toroidal current density and q-profile at the nonlinear stage of (m = 1/n = 1) kink instability of a tokamak plasma is presented. The flattening is shown to be an essentially multi-scale phenomenon which is characterized by, at least, two magnetic Reynolds numbers. The ordinary one, R m , is related with a characteristic radial scale-length, while the other, R m * , corresponds to a characteristic scale-length of plasma inhomogeneity along the magnetic field line. In a highly conducting plasma inside the q = 1 magnetic surface, where q value does not much differ from unity, plasma evolution is governed by a multi-scale non-ideal dynamics characterized by two well-separated magnetic Reynolds numbers, R m and R m * ≡ (1 - q) R m , where R m * - O(1) and R m >> 1. This dynamics consistently explains two seemingly contradictory features recently observed in a numerical simulation [Watanabe et al., 1995]: i) the current profile (q-profile) is flattened in the magnetohydrodynamic time scale within the q = 1 rational surface; ii) the magnetic surface keeps its initial circular shape during this evolution. (author)

  3. Magnetic ''islandography'' in tokamaks

    International Nuclear Information System (INIS)

    Callen, J.D.; Waddell, B.V.; Hicks, H.R.

    1978-09-01

    Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

  4. Energy confinement and transport of H-mode plasmas in tokamak

    International Nuclear Information System (INIS)

    Urano, Hajime

    2005-02-01

    A characteristic feature of the high-confinement (H-mode) regime is the formation of a transport barrier near the plasma edge, where steepening of the density and temperature gradients is observed. The H-mode is expected to be a standard operation mode in a next-step fusion experimental reactor, called ITER-the International Thermonuclear Experimental Reactor. However, energy confinement in the H-mode has been observed to degrade with increasing density. This is a critical constraint for the operation domain in the ITER. Investigation of the main cause of confinement degradation is an urgent issue in the ITER Physics Research and Development Activity. A key element for solving this problem is investigation of the energy confinement and transport properties of H-mode plasmas. However, the influence of the plasma boundary characterized by the transport barrier in H-modes on the energy transport of the plasma core has not been examined sufficiently in tokamak research. The aim of this study is therefore to investigate the energy confinement properties of H-modes in a variety of density, plasma shape, seed impurity concentration, and conductive heat flux in the plasma core using the experimental results obtained in the JT-60U tokamak of Japan Atomic Energy Research Institute. Comparison of the H-mode confinement properties with those of other tokamaks using an international multi-machine database for extrapolation to the next step device was also one of the main subjects in this study. Density dependence of the energy confinement properties has been examined systematically by separating the thermal stored energy into the H-mode pedestal component determined by MHD stability called the Edge Localized Modes (ELMs) and the core component governed by gyro-Bohm-like transport. It has been found that the pedestal pressure imposed by the destabilization of ELM activities led to a reduction in the pedestal temperature with increasing density. The core temperature for each

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

    International Nuclear Information System (INIS)

    Kurihara, Kenichi; Kimura, Toyoaki

    1992-05-01

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

  6. Thermonuclear-driven fast magnetosonic-wave heating in tokamak plasmas

    International Nuclear Information System (INIS)

    Sutton, W.R. III.

    1982-01-01

    A thermonuclear driven fast magnetosonic wave instability is investigated in tokamak plasmas for propagation transverse to the external magnetic field at frequencies of several times the alpha particle gyro rate: ω approx. = L Ω/sub α/ = k/sub perpendicular/ v/sub A/, L approx. 4 to 8, k/sub parallel/ << k/sub perpendicular/. The 2-D differential quasi-linear diffusion equation is derived in circular cylindrical, v/sub perpendicular/-v/sub parallel/ geometry. We perform an expansion in the small parameter k/sub parallel/k/sub perpendicucular/ of the quasi-linear diffusion coefficients

  7. Microwave reflectrometry for electron density measurements in the TJ-1 tokamak plasma

    International Nuclear Information System (INIS)

    Anabitarte, E.; Bustamante, E.G.; Calderon, M.A.G.; Vegas, A.

    1986-01-01

    A study about microwave reflectometry to measure the outside profile of the electron plasma density on tokamak TJ-1 is presented. It is also presented the condition of applicability of this method after the characteristic parameters of the plasma and its resolution. The simulation of the plasma in laboratory by means of a metallic mirror causes the whole characterization of the reflectometer. (author)

  8. A study of quasi-mode parametric excitations in lower-hybrid heating of tokamak plasmas

    International Nuclear Information System (INIS)

    Villalon, E.; Bers, A.

    1980-01-01

    A detailed linear and non-linear analysis of quasi-mode parametric excitations relevant to experiments in supplementary heating of tokamak plasmas is presented. The linear analysis includes the full ion-cyclotron harmonic quasi-mode spectrum. The non-linear analysis, considering depletion of the pump electric field, is applied to the recent Alcator A heating experiment. Because of the very different characteristics of a tokamak plasma near the wall (in the shadow of the limiter) and inside, the quasi-mode excitations are studied independently for the plasma edge and the main bulk of the plasma, and for two typical regimes in overall density, the low (peak in density, n 0 =1.5x10 14 cm -3 ) and high (n 0 =5x10 14 cm -3 ) density regimes. At the edge of the plasma and for the low-density regime, it is found that higher nsub(z)(nsub(z)=cksub(z)/ω) than those predicted by the linear theory are strongly excited. Inside the plasma, the excitation of higher wave numbers is also significant. These results indicate that a large amount of the RF-power may not penetrate to the plasma centre, but will rather be either Landau-damped on the electrons or mode-converted into thermal modes, close to the plasma edge. Moreover, for sufficiently high peaks in density, it is found that all the RF-power is mode-converted before reaching the plasma centre. Inside the plasma, the power density of the excited sideband fields is shown to be always very small in comparison with their excitation at the plasma edge. (author)

  9. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

  10. Tokamak ARC damage

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  11. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

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

  12. The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Tammen, H.F.

    1995-01-10

    One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics `Rijnhuizen`, was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL).

  13. The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

    International Nuclear Information System (INIS)

    Tammen, H.F.

    1995-01-01

    One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics 'Rijnhuizen', was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL)

  14. The effect of tangled magnetic fields on instabilities in tokamak plasmas

    International Nuclear Information System (INIS)

    Thornton, A J; Kirk, A; Harrison, J R; Chapman, I T; Cahyna, P; Nardon, E

    2014-01-01

    The high pressure gradients in the edge of a tokamak plasma can lead to the formation of explosive plasma instabilities known as edge localised modes (ELMs). The control of ELMs is an important requirement for the next generation of fusion devices such as ITER. Experiments performed on the Mega Amp Spherical Tokamak (MAST) at Culham have shown that the application of non-axisymetric resonant magnetic perturbations (RMPs) can be used to mitigate ELMs. During the application of the RMPs, clear structures are observed in visible- light imaging of the X-point region. These lobes, or tangles, have been observed for the first time and their appearance is correlated with the mitigation of ELMs. Tangle formation is seen to be associated with the RMPs penetrating the plasma and may be important in explaining why the ELM frequency increases during ELM mitigation. Whilst the number and location of the tangles can be explained by vacuum magnetic field modelling, obtaining the correct radial extent of the tangles requires the plasma response to be taken into account

  15. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  16. Dynamic behavior of plasma-facing materials during plasma instabilities in tokamak reactors

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1997-01-01

    Damage to plasma-facing and nearby components due to plasma instabilities remains a major obstacle to a successful tokamak concept. The high energy deposited on facing materials during plasma instabilities can cause severe erosion, plasma contamination, and structural failure of these components. Erosion damage can take various forms such as surface vaporization, spallation, and liquid ejection of metallic materials. Comprehensive thermodynamic and radiation hydrodynamic codes have been developed, integrated, and used to evaluate the extent of various damage to plasma-facing and nearby components. The eroded and splashed materials will be transported and then redeposited elsewhere on other plasma-facing components. Detailed physics of plasma/solid-liquid/vapor interaction in a strong magnetic field have been developed, optimized, and implemented in a self-consistent model. The plasma energy deposited in the evolving divertor debris is quickly and intensely reradiated, which may cause severe erosion and melting of other nearby components. Factors that influence and reduce vapor-shielding efficiency such as vapor diffusion and turbulence are also discussed and evaluated

  17. Neoclassical MHD descriptions of tokamak plasmas

    International Nuclear Information System (INIS)

    Callen, J.D.; Kim, Y.B.; Sundaram, A.K.

    1988-01-01

    Considerable progress has been made in extending neoclassical MHD theory and in exploring the linear instabilities, nonlinear behavior and turbulence models it implies for tokamak plasmas. The areas highlighted in this paper include: extension of the neoclassical MHD equations to include temperature-gradient and heat flow effects; the free energy and entropy evolution implied by this more complete description; a proper ballooning mode formalism analysis of the linear instabilities; a new rippling mode type instability; numerical simulation of the linear instabilities which exhibit a smooth transition from resistive ballooning modes at high collisionality to neoclassical MHD modes at low collisionality; numerical simulation of the nonlinear growth of a single helicity tearing mode; and a Direct-Interaction-Approximation model of neoclassical MHD turbulence and the anomalous transport it induces which substantially improves upon previous mixing length model estimates. 34 refs., 2 figs

  18. Role of boundary plasma in lower-hybrid-frequency heating of a tokamak

    International Nuclear Information System (INIS)

    Uehara, Kazuya; Yamamoto, Takumi; Fujii, Tsuneyuki

    1982-01-01

    Boundary plasma of a circular tokamak has been investigated by means of electrostatic probes during lower-hybrid heating. The reflection coefficient is affected by the density gradient in front of the launcher. An effective ion heating is performed in the main plasma region when the boundary electron temperature is relatively high enough to suppress the parametric decay instabilities. The simultaneous injection of neutral beams as well as the lower-hybrid wave brings the suppression of instabilities with increase of the electron temperature coming from the neutral beam heating. (author)

  19. 2-D Imaging of Electron Temperature in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Munsat, T.; Mazzucato, E.; Park, H.; Domier, C.W.; Johnson, M.; Luhmann, N.C. Jr.; Wang, J.; Xia, Z.; Classen, I.G.J.; Donne, A.J.H.; Pol, M.J. van de

    2004-01-01

    By taking advantage of recent developments in millimeter wave imaging technology, an Electron Cyclotron Emission Imaging (ECEI) instrument, capable of simultaneously measuring 128 channels of localized electron temperature over a 2-D map in the poloidal plane, has been developed for the TEXTOR tokamak. Data from the new instrument, detailing the MHD activity associated with a sawtooth crash, is presented

  20. Frequency spectral broadening of lower hybrid waves in tokamak plasmas - causes and effects

    Energy Technology Data Exchange (ETDEWEB)

    Pericoli Ridolfini, V; Giannone, L.; Bartiromo, R [Associazione Euratom-ENEA sulla Fusione, Rome (Italy). Centro Ricerche Energia Frascati

    1994-04-01

    The frequency spectral broadening of lower hybrid (LH) waves injected into tokamak plasmas is extensively analyzed with reference mostly to experimental data from the ASDEX tokamak. The link between the magnitude of the pump spectral width and the degradation of the LH current drive efficiency (up to a factor of 2), pointed out in previous works, is explained. The experimental behaviour of LH power absorption is also well reproduced, even in situations when the access of the launched LH waves to the core plasma should be largely forbidden. Experiments are described that are aimed at determined whether parametric decay instabilities (PDIs) or scattering of LH waves by density fluctuations in the plasma edge are causes of the broadening of the LH pump frequency spectrum. Fluctuations emerge as the largely dominant process, while no signature of PDI processes is observed. Careful measurements of the density fluctuations in the ASDEX scrape-off layer plasma allow the analytical description given by Andrews and Perkins to be assumed as the appropriate model of LH scattering. Indeed, it supplies the correct magnitude for the frequency spectral width of the LH pump, and explains quantitatively, together with a ray tracing code, why the CD efficiency decreases with the broadening of the pump spectrum. It can also account for the observed LH power absorption coefficient. (author). 48 refs, 13 figs, 2 tabs.

  1. Fusion reactivity, confinement, and stability of neutral-beam heated plasmas in TFTR and other tokamaks

    International Nuclear Information System (INIS)

    Park, Hyeon, K.

    1996-05-01

    The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

  2. The rate of plasma heating by harmonic ion cyclotron waves in tokamaks

    International Nuclear Information System (INIS)

    Moslehi-Fard, M.; Sobhanian, S.; Solati-Kia, F.

    2002-01-01

    In tokamaks, the toroidal magnetic field, B φ , is due to the current in coils around plasma, and the poloidal magnetic field B p results from the plasma itself. Usually B φ p , and the combination of these two fields forms a nested set of toroidal magnetic surfaces. The equilibrium Grad-Shafranov equation is investigated and it is shown that the particle products of fusion with different pitch angles on these surfaces have different orbital shapes. In the JET tokamak, the α particles with pitch angle θ smaller than 54.8 deg are passing, those with θ between 54.8 deg and 65.1 deg have trapping-passing orbits but for θ greater than 65.1 deg the orbit has a banana form. Other tokamaks such as Alcator and ITER are also considered. The passing, trapping-passing and banana orbits in these tokamaks are traced. The results obtained from this calculation are analyzed. The wave damping has been investigated produced from interaction with particles, particularly α particles, and the rate of heating for l = 1 to 8 harmonics is plotted. The results of calculation show that heating at the fourth harmonic reaches a maximum. For higher harmonics, the heating does not change much from the fourth harmonic. (author)

  3. Wave trajectory and electron cyclotron heating in tokamak plasmas

    International Nuclear Information System (INIS)

    Tanaka, S.; Maekawa, T.; Terumichi, Y.; Hamada, Y.

    1980-01-01

    Wave trajectories in high density tokamak plasmas are studied numerically. Results show that the ordinary wave injected at an appropriate incident angle can propagate into the dense plasmas and is mode-converted to the extraordinary wave at the plasma cutoff, is further converted to the electron Bernstein wave during passing a loop or a folded curve near the upper hybrid resonance layer, and is cyclotron damped away, resulting in local electron heating before arriving at the cyclotron resonance layer. Similar trajectory and damping are obtained when a microwave in a form of extraordinary wave is injected quasi-perpendicularly in the direction of decreasing toroidal field

  4. Scattering effects of small-scale density fluctuations on reflectometric measurements in a tokamak plasma

    International Nuclear Information System (INIS)

    Garcia, J.P.; Manso, M.E.; Serra, F.M.; Mendonca, J.T.

    1989-01-01

    When a wave propagates in a non homogeneous fluctuating plasma part of the incident energy is scattered out to the nonlinear interaction between the wave and the oscillating modes perturbing the plasma. The possibility of enhanced scattering at the cutoff layer, where reflection of the incident wave occurs, has been recently suggested as the basis of a reflectometric experiment to determine the spatial location of small scale fluctuations in a fusion plasma. Here we report on the development of a theoretical model to evaluate the flux of energy scattered by fluctuations, in order to give insight about the interpretation of measurements using a microwave reflectometry diagnostic in a tokamak. The scattered field is obtained through the resolution of a (non-homogeneous) wave propagation equation where the source term is related with the nonlinear current due to the interaction between the incident wave and local fluctuations. We use a slab model for the plasma, and an ordinary (0) wave propagation along the density gradient is considered. The amplitude of the scattered wave at the border of the plasma is estimated. In order to know the contributions to the energy scattered both from the propagation region and the reflecting layer, an approach was used where perturbations are modelled by spatial step functions at several layers. The main contribution to the scattered power comes from the cutoff region, where the electric field amplitude swells as compared with the incident value. Considering the reflectometric system recently installed on the ASDEX tokamak, and using typical density profiles, expected values of the 'swelling factor' have been numerically evaluated. The role of incoherent scattering due to drift wave activity is discussed as well as the coherent scattering due to fluctuations induced by lower hybrid (LH) waves. (author) 2 refs., 4 figs

  5. Study on possibility of plasma current profile determination using an analytical model of tokamak equilibrium

    International Nuclear Information System (INIS)

    Moriyama, Shin-ichi; Hiraki, Naoji

    1996-01-01

    The possibility of determining the current profile of tokamak plasma from the external magnetic measurements alone is investigated using an analytical model of tokamak equilibrium. The model, which is based on an approximate solution of the Grad-Shafranov equation, can set a plasma current profile expressed with four free parameters of the total plasma current, the poloidal beta, the plasma internal inductance and the axial safety factor. The analysis done with this model indicates that, for a D-shaped plasma, the boundary poloidal magnetic field prescribing the external magnetic field distribution is dependent on the axial safety factor in spite of keeping the boundary safety factor and the plasma internal inductance constant. This suggests that the plasma current profile is reversely determined from the external magnetic analysis. The possibility and the limitation of current profile determination are discussed through this analytical result. (author)

  6. The simulation of L-H transition in tokamak plasma using MMM95 transport model

    International Nuclear Information System (INIS)

    Intharat, P; Poolyarat, N; Chatthong, B; Onjun, T; Picha, R

    2015-01-01

    BALDUR integrative predictive modelling code together with a Multimode (MMM95) anomalous transport model is used to simulate the evolution profiles, including plasma current, temperature, density and energy in a tokamak reactor. It is found that a self - transition from low confinement mode (L-mode) to high confinement mode (H-mode) regimes can be achieved once a sufficient auxiliary heating applied to the plasma is reached. The result agrees with experimental observations from various tokamaks. A strong reduction of turbulent transport near the edge of plasma is also observed, which is related to the formation of steep radial electric field near the edge regime. From transport analysis, it appears that the resistive ballooning mode is the dominant term near the plasma edge regime, which is significantly reduced during the transition. (paper)

  7. Ion temperature measurements of H-, D- and He-plasmas in the TCA tokamak by collective Thomson scattering of D2O laser radiation

    International Nuclear Information System (INIS)

    Behn, R.; Dicken, D.; Hackmann, J.; Salito, S.A.; Siegrist, M.R.

    1989-01-01

    Development of collective Thomson scattering as a method to measure the ion temperature of a tokamak plasma has been successful and encouraging results have been obtained during experiments on TCA in H-, D- and He-plasmas. Using a laser source in the far-infrared spectral region allows scattering angles close to 90 o , which results in excellent spatial resolution. The system installed on the TCA tokamak comprises an optically pumped D 2 O laser emitting 0.5 J in a 1.4 μs pulse on its Raman transition at 385μm. A heterodyne receiver with a Schottky barrier diode mixer has been chosen to detect the scattered radiation and analyze its spectral distribution in 12 channels of 80 MHz. Recent improvements of the mixer and 1st IF-amplifier yielded a system NEP of 2.2·10 -19 W/Hz. As a consequence we have obtained results which allow for the first time to evaluate the ion temperature T i in a single laser shot. (author) 3 figs., 1 tab

  8. Positional instability analysis of tokamak plasmas by ERATO

    International Nuclear Information System (INIS)

    Kumagai, Michikazu; Tsunematsu, Toshihide; Tokuda, Shinji; Takeda, Tatsuoki

    1983-06-01

    The stability of axisymmetric modes of a tokamak plasma(positional instabilities) is analyzed for the Solov'ev equilibrium by using the linear ideal MHD code ERATO-J. The dependence of the stability on various parameters, i.e., the ellipticity and triangularity of the plasma cross-section, the aspect ratio, the safety factor at the magnetic axis, and the distance between the plasma and a conducting shell is investigated. Comparison of the results with those by the rigid model shows that the stability condition derived from the rigid model in terms of the decay index(n-index) of the external equilibrating field is a good approximation for the plasma with small triangular deformation. Also the results are compared with those of the rigid displacement model and applicability of the various models on the positional instability analyses is discussed. (author)

  9. Alignment and experiment of the HL-1 X-ray bent-crystal spectrometer

    International Nuclear Information System (INIS)

    Chen Jiefu; Yuan Chengji; Zhang Shuxun; Lu Jie; Yie Gaoying; Zhang Li

    1993-04-01

    An X-ray bent-crystal spectrometer, which is developed by Southwestern Institute of Physics, has been aligned and experimental on the HL-1 Tokamak device. It has been used to acquire experimental spectrum. This spectrometer has time resolution function and is a high through-put and high resolution Bragg crystal spectrometer with Johann configuration. It uses a large quartz lamina, its effective area is 11 x 5 cm 2 , with a radius of curvature of 377 cm as the dispersion element. The detector is a large size (10 x 10 cm 2 ) one-dimensional resolving multiwire proportional counter, and the spectral resolving power is R ≅ 18000. Under the injection of Ar into hydrogen plasma, the He-like Ar ion satellite spectra at some wavelength ranges are obtained. The central ion temperature has been given out from the Doppler broadening of He-like Ar x VII 3.9457 angstrom resonance line. This shows the success of spectrometer developed. The experimental results and the further improvement of this spectrometer are also discussed

  10. Inertia effects on the rigid displacement approximation of tokamak plasma vertical motion

    International Nuclear Information System (INIS)

    Carrera, R.; Khayrutdinov, R.R.; Azizov, E.A.; Montalvo, E.; Dong, J.Q.

    1991-01-01

    Elongated plasmas in tokamaks are unstable to axisymmetric vertical displacements. The vacuum vessel and passive conductors can stabilize the plasma motion in the short time scale. For stabilization of the plasma movement in the long time scale an active feedback control system is required. A widely used method of plasma stability analysis uses the Rigid Displacement Model (RDM) of plasma behavior. In the RDM it is assumed that the plasma displacement is small and usually plasma inertia effects are neglected. In addition, it is considered that no changes in plasma shape, plasma current, and plasma current profile take place throughout the plasma motion. It has been demonstrated that the massless-filament approximation (instantaneous force-balance) accurately reproduces the unstable root of the passive stabilization problem. Then, on the basis that the instantaneous force-balance approximation is correct in the passive stabilization analysis, the massless approximation is utilized also in the study of the plasma vertical stabilization by active feedback. The authors show here that the RDM (without mass effects included) does not provide correct stability results for a tokamak configuration (plasma column, passive conductors, and feedback control coils). Therefore, it is concluded that inertia effects have to be retained in the RDM system of equations. It is shown analytically and numerically that stability diagrams with and without plasma-mass corrections differ significantly. When inertia effects are included, the stability region is more restricted than obtained in the massless approximation

  11. Wall conditioning of the TBR-1 Tokamak by plasma generated by microwaves

    International Nuclear Information System (INIS)

    Elizondo, J.I.

    1986-01-01

    A new system of vaccum chamber wall conditioning in the TBR-1 Tokamak, using electron cyclotron resonance plasma of hydrogen for the discharge cleaning process is presented. The construction and performance of equipments are described, and the cleaning process to otimize the conditioning efficiency by chase of plasma parameters. (author) [pt

  12. Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

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

  13. Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

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

  14. Predictive modelling of edge transport phenomena in ELMy H-mode tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Loennroth, J.-S.

    2009-01-01

    This thesis discusses a range of work dealing with edge plasma transport in magnetically confined fusion plasmas by means of predictive transport modelling, a technique in which qualitative predictions and explanations are sought by running transport codes equipped with models for plasma transport and other relevant phenomena. The focus is on high confinement mode (H-mode) tokamak plasmas, which feature improved performance thanks to the formation of an edge transport barrier. H-mode plasmas are generally characterized by the occurrence of edge localized modes (ELMs), periodic eruptions of particles and energy, which limit confinement and may turn out to be seriously damaging in future tokamaks. The thesis introduces schemes and models for qualitative study of the ELM phenomenon in predictive transport modelling. It aims to shed new light on the dynamics of ELMs using these models. It tries to explain various experimental observations related to the performance and ELM-behaviour of H-mode plasmas. Finally, it also tries to establish more generally the potential effects of ripple-induced thermal ion losses on H-mode plasma performance and ELMs. It is demonstrated that the proposed ELM modelling schemes can qualitatively reproduce the experimental dynamics of a number of ELM regimes. Using a theory-motivated ELM model based on a linear instability model, the dynamics of combined ballooning-peeling mode ELMs is studied. It is shown that the ELMs are most often triggered by a ballooning mode instability, which renders the plasma peeling mode unstable, causing the ELM to continue in a peeling mode phase. Understanding the dynamics of ELMs will be a key issue when it comes to controlling and mitigating the ELMs in future large tokamaks. By means of integrated modelling, it is shown that an experimentally observed increase in the ELM frequency and deterioration of plasma confinement triggered by external neutral gas puffing might be due to a transition from the second to

  15. Calculation of the neoclassical conductivity of plasma and fraction of trapped particles for elongated Damavand Tokamak

    International Nuclear Information System (INIS)

    Dini, F.; Khorasani, S.

    2007-01-01

    Configuration of Tokamak plasma has a dominant effect on its parameters. In the calculation of transport, there are some transport coefficients and quantities, where the knowledge of their precise values, according to the system of equations, is essential to be realized. Tokamak has a toroidal configuration, in addition to classical effects, it is necessary to study the neoclassical effects due to the field curvature. The trapped particles in strong electromagnetic fields oscillate on banana-shaped orbits which in turn affect many other collisional transport parameters. Here, a precise estimation of trapped particles based on the standard equilibrium model for an elliptical shape of Tokamak plasma has been carried out using Lin-Liu model. It should be added that in this calculation, the profile of the averaged magnetic field on the flux surfaces has been derived using analytical integration and consideration of an elliptic shape for ellipticity function in the limit of large aspect ratio and zero shift of magnetic flux surfaces. Having the fraction of the trapped particles, by ,following the formulation and using an appropriate model in various collisional regimes, the neoclassical conductivity of plasma in Damavand Tokamak is obtained and the respective variations have been found. The presented results can exploit the computation of transport and other quantities of Damavand Tokamak

  16. Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Anda, G.; Bencze, A.; Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Háček, P., E-mail: hacek@ipp.cas.cz [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Hron, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Kovácsik, A. [Wigner – RCP, HAS, Budapest (Hungary); Department of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); 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); Pánek, R. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Réfy, D.; Veres, G. [Wigner – RCP, HAS, Budapest (Hungary); Weinzettl, V. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

    2015-10-15

    Highlights: • Li-BES detection system on the COMPASS tokamak is optimized observation system with high temporal resolution. • High sensitivity to low level light fluctuations. • Optics and detectors with electronics are placed in thermally stabilized compact box. • Fast deflection system allows us to measure background corrected electron density profiles on microsecond time-scale. - Abstract: A new Li beam emission spectroscopy (Li-BES) diagnostic system with a ∼ cm spatial resolution, and with beam energy ranging from 10 keV up to 120 keV and a 18 channel Avalanche photo diode (APD) detector system sampled at 2 MHz has been recently installed and tested on the COMPASS tokamak. This diagnostic allows to reconstruct density profile based on directly measured light profiles, and to follow turbulent behaviour of the edge plasma. The paper reports technical capabilities of this new system designed for fine spatio-temporal measurements of plasma electron density. Focusing on turbulence-induced fluctuation measurements, we demonstrate how physically relevant information can be extracted using the COMPASS Li-BES system.

  17. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    Hendel, H.W.; Jassby, D.L.

    1989-11-01

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  18. Pellet-plasma interaction: Local disturbances caused by pellet ablation in tokamaks

    International Nuclear Information System (INIS)

    Lengyel, L.L.

    1989-01-01

    The local disturbance amplitudes caused by ablating pellets in tokamaks are computed in the framework of a magnetohydrodynamic model supplemented by the neutral gas plasma shielding ablation model. The model computes, for a given number of pellet particles locally deposited, the time histories of the ablatant cloud parameters, such as cloud radius, cloud length, electron density, temperature and cloud beta, at a succession of magnetic flux surfaces. The cloud radius thus determined may be fed back into the ablation model, thus adjusting the effect of the shielding cloud on the ablation rate. The model is applied to typical plasma parameter ranges of existing and planned tokamaks. The results show that the ablating pellets may cause massive local disturbances in tokamaks, depending upon the number of particles locally deposited. The peaks of these disturbances are of a spike nature, lasting only a few microseconds (Alfven time-scale). The characteristic decay time of the 'quasi-steady' disturbance values that characterize the after-spike period is of the order of several milliseconds (hydrodynamic time-scale). The peak electron density values may be as high as 10 23 to 10 25 m -3 , with the associated beta peaks exceeding unity. The 'quasi-steady' values of the electron density and the ablatant beta may be of the order of 10 22 to 10 24 m -3 and unity, respectively. Furthermore, the results show the strong dependence of the ablation rate on the dynamic characteristics of the ablatant cloud surrounding the pellet. (author). 25 refs, 6 figs, 2 tabs

  19. High beta plasmas in the PBX tokamak

    International Nuclear Information System (INIS)

    Bol, K.; Buchenauer, D.; Chance, M.

    1986-04-01

    Bean-shaped configurations favorable for high β discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present β limit

  20. Multi-scale semi-ideal magnetohydrodynamics of a tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Bazdenkov, S.; Sato, Tetsuya; Watanabe, Kunihiko

    1995-09-01

    An analytical model of fast spatial flattening of the toroidal current density and q-profile at the nonlinear stage of (m = 1/n = 1) kink instability of a tokamak plasma is presented. The flattening is shown to be an essentially multi-scale phenomenon which is characterized by, at least, two magnetic Reynolds numbers. The ordinary one, R{sub m}, is related with a characteristic radial scale-length, while the other, R{sub m}{sup *}, corresponds to a characteristic scale-length of plasma inhomogeneity along the magnetic field line. In a highly conducting plasma inside the q = 1 magnetic surface, where q value does not much differ from unity, plasma evolution is governed by a multi-scale non-ideal dynamics characterized by two well-separated magnetic Reynolds numbers, R{sub m} and R{sub m}{sup *} {identical_to} (1 - q) R{sub m}, where R{sub m}{sup *} - O(1) and R{sub m} >> 1. This dynamics consistently explains two seemingly contradictory features recently observed in a numerical simulation [Watanabe et al., 1995]: (i) the current profile (q-profile) is flattened in the magnetohydrodynamic time scale within the q = 1 rational surface; (ii) the magnetic surface keeps its initial circular shape during this evolution. (author).

  1. Numerical studies of transport processes in Tokamak plasma

    International Nuclear Information System (INIS)

    Spineanu, F.; Vlad, M.

    1984-09-01

    The paper contains the summary of a set of studies of the transport processes in tokamak plasma, performed with a one-dimensional computer code. The various transport models (which are implemented by the expressions of the transport coefficients) are presented in connection with the regimes of the dynamical development of the discharge. Results of studies concerning the skin effect and the large scale MHD instabilities are also included

  2. First results from plasma density measurements in the FTU tokamak by means of a two-frequency pulsed time-of-flight refractometer

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, V. G.; Malyshev, A. Yu.; Markov, V. K.; Petrov, A. A. [Troitsk Institute for Innovation and Fusion Research (Russian Federation); Avino, F.; Angelis, R. de; Tudisco, O. [ENEA-UT Fusione Centro Ricerche Frascati (Italy)

    2012-04-15

    A pulsed time-of-flight refractometer was developed and tested to determine the mean plasma density in the T-11M tokamak by measuring the propagation time of nanosecond microwave pulses in plasma. Later, it was also proposed to use such an instrument to measure and control the mean plasma density in the ITER tokamak by probing the plasma with an extraordinary wave, the electric field of which is perpendicular to the magnetic field in plasma, in the transparency window at frequencies of 50-100 GHz. To avoid the effect of the density profile shape on the measurement results in the nonlinear mode of refractometer operation (near the cutoff), a system operating at two different probing frequencies was developed and tested. Such a system provides two values of the time delay, which can be used to estimate the peaking factor of the density distribution {alpha} and correctly determine the linear density Left-Pointing-Angle-Bracket Nl Right-Pointing-Angle-Bracket , regardless of the density profile (assuming a smooth density profile of the form of N({rho}) = N(0)(1 - {rho}{sup 2}){sup {alpha}}, where N(0) is the central plasma density and {rho} = r/a is the normalized plasma radius). The first experiments on density measurements in the FTU tokamak performed with this refractometer are described, and results from these experiments are presented. The formation of a thin dense plasma layer in the zone of a strong magnetic field (the so-called MARFE layer) at a relatively low (for FTU) plasma density of {approx}6 Multiplication-Sign 10{sup 19} m{sup -3} was detected. The thickness of this layer, determined from the refractometry data, agrees well with the data obtained using a digital camera.

  3. Sawtooth Pacing by Real-Time Auxiliary Power Control in a Tokamak Plasma

    International Nuclear Information System (INIS)

    Goodman, T. P.; Felici, F.; Sauter, O.; Graves, J. P.

    2011-01-01

    In the standard scenario of tokamak plasma operation, sawtooth crashes are the main perturbations that can trigger performance-degrading, and potentially disruption-generating, neoclassical tearing modes. This Letter demonstrates sawtooth pacing by real-time control of the auxiliary power. It is shown that the sawtooth crash takes place in a reproducible manner shortly after the removal of that power, and this can be used to precisely prescribe, i.e., pace, the individual sawteeth. In combination with preemptive stabilization of the neoclassical tearing modes, sawtooth pacing provides a new sawtooth control paradigm for improved performance in burning plasmas.

  4. A method for measuring plasma position in the rectangular tokamak TJ-I

    International Nuclear Information System (INIS)

    Qin, J.; Ascasibar, E.; Pastor, I.; Navarro, A.P.; Ochando, M.A.; Pedrosa, M.A.; Rodriguez, L.; Sanchez, J.

    1993-01-01

    Determinations of plasma position in tokamaks are based on measuring the external magnetic field of the plasma current. Usually, the measurements are performed with magnetic coils along a closed contour around the plasma column, or with small magnetic probes in pairs positioned diametrically. However, such measurements are constrained in TJ-I because of a severe restriction on installing the magnetic probes. It is only possible to install some small magnetic probes inside the vacuum vessel near its top and bottom walls where the limiters stand. In this paper an alternative method of measuring plasma position in TJ-I is presented. The plasma position can be derived from the measurements of those magnetic probes located off the equatorial plane, with the aid of a filament model. The method is proved to be reliable, and would be applied to those small-size devices with similar restrictions. (author) 5 refs., 3 figs., 1 tab

  5. Effect of ECRH and resonant magnetic fields on formation of magnetic islands in the T-10 tokamak plasma

    Science.gov (United States)

    Shestakov, E. A.; Savrukhin, P. V.

    2017-10-01

    Experiments in the T-10 tokamak demonstrated possibility of controlling the plasma current during disruption instability using the electron cyclotron resonance heating (ECRH) and the controlled operation of the ohmic current-holding system. Quasistable plasma discharge with repeating sawtooth oscillations can be restored after energy quench using auxiliary ECRH power when PEC / POH > 2-5. The external magnetic field generation system consisted of eight saddle coils that were arranged symmetrically relative to the equatorial plane of the torus outside of the vacuum vessel of the T-10 tokamak to study the possible resonant magnetic field effects on the rotation frequency of magnetic islands. The saddle coils power supply system is based on four thyristor converters with a total power of 300 kW. The power supply control system is based on Siemens S7 controllers. As shown by preliminary experiments, the interaction efficiency of external magnetic fields with plasma depends on the plasma magnetic configuration. Optimal conditions for slowing the rotation of magnetic islands were determined. Additionally, the direction of the error magnetic field in the T-10 tokamak was determined, and the threshold value of the external magnetic field was determined.

  6. Plasma jet source parameter optimisation and experiments on injection into Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Petrov, Yu.V.; Sakharov, N.V.; Semenov, A.A.; Voronin, A.V.

    2005-01-01

    Results of theoretical and experimental research on the plasma sources and injection of plasma and gas jet produced by the modified source into tokamak Globus-M are presented. An experimental test stand was developed for investigation of intense plasma jet generation. Optimisation of pulsed coaxial accelerator parameters by means of analytical calculations is performed with the aim of achieving the highest flow velocity at limited coaxial electrode length and discharge current. The optimal parameters of power supply to generate a plasma jet with minimal impurity contamination and maximum flow velocity were determined. A comparison of experimental and calculation results is made. Plasma jet parameters are measured, such as: impurity species content, pressure distribution across the jet, flow velocity, plasma density, etc. Experiments on the interaction of a higher kinetic energy plasma jet with the magnetic field and plasma of the Globus-M tokamak were performed. Experimental results on plasma and gas jet injection into different Globus-M discharge phases are presented and discussed. Results are presented on the investigation of plasma jet injection as the source for discharge breakdown, plasma current startup and initial density rise. (author)

  7. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    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

  8. The O-X-B mode conversion scheme for ECRH of a high-density Tokamak plasma

    DEFF Research Database (Denmark)

    Hansen, F. R.; Lynov, Jens-Peter; Michelsen, Poul

    1985-01-01

    A method to apply electron cyclotron resonance heating (ECRH) to a Tokamak plasma with central density higher than the critical density for cut-off of the ordinary mode (O-mode) has been investigated. This method involves two mode conversions, from an O-mode via an extraordinary mode (X......-mode) into an electron Bernstein mode (B-mode). Radial profiles for the power deposition and the wave-drive current due to the B-waves are calculated for realistic antenna radiation patterns with parameters corresponding to the Danish DANTE Tokamak and to Princeton's PLT....

  9. Conceptual design and simulation of HL-2A RMP coils power supply

    International Nuclear Information System (INIS)

    Ren Qinghua; Wang Yingqiao; Li Qing

    2015-01-01

    The conceptual design and simulation of HL-2A resonant magnet perturbation (RMP) coils power supply are introduced, including its power supply topology, the selection of basic units and its control modes. The RMP coils will be energized by DC power supply with the amplitude up to 4 kA, or AC power supply with maximum amplitude up to 4 kA with adjustable frequency up to 1 kHz. Though choosing the sound power supply topology and its control modes, the RMP coils power supply can meet the requirements of both fast response and low current ripple to satisfy the ELM control and minimize adverse effects on plasma control as far as possible. (authors)

  10. A study on the fusion reactor - Development of MHD stability and transport code for KT-2 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Koo; Shin, Kyo Jin [Pohang University of Science and Tecnology, Pohang (Korea, Republic of)

    1996-08-01

    MHD Stability analyses for KT-2 Tokamak were carried out by using CART (Resistive 3-D) Code. Linear Growth rates and linear perturbed eigen function of both N=0 axisymmetric mode and N=1 kink modes of highly elongated tokamak plasmas, in the presence of a conducting wall at various distances are computed and linear and nonlinear evolution of N=0 axisymmetric modes are simulated. 26 refs., 25 figs. (author)

  11. Magnetohydrodynamic helical structures in nominally axisymmetric low-shear tokamak plasmas

    International Nuclear Information System (INIS)

    Graves, J P; Brunetti, D; Cooper, W A; Reimerdes, H; Halpern, F; Pochelon, A; Sauter, O; Chapman, I T

    2013-01-01

    The primary goal of hybrid scenarios in tokamaks is to enable high performance operation with large plasma currents whilst avoiding MHD instabilities. However, if a local minimum in the safety factor is allowed to approach unity, the energy required to overcome stabilizing magnetic field line bending is very small, and as a consequence, large MHD structures can be created, with typically dominant m = n = 1 helical component. If there is no exact q = 1 rational surface the essential character of these modes can be modelled assuming ideal nested magnetic flux surfaces. The methods used to characterize these structures include linear and non-linear ideal MHD stability calculations which evaluate the departure from an axisymmetric plasma state, and also equilibrium calculations using a 3D equilibrium code. While these approaches agree favourably for simulations of ITER relevant hybrid regimes in this paper, the relevance of the ideal MHD model itself is tested through empirical examination of helical states in MAST and TCV. While long lived modes in MAST do not have island structures, some of the continuous mode oscillations exhibited in high elongation experiments in TCV indicate that resistivity may play a role in further weakening the ability of the tokamak core to remain axisymmetric. The simulations and experiments consistently highlight the need to control the safety factor in hybrid scenarios planned for future fusion grade tokamaks such as ITER. (paper)

  12. The use of internal transport barriers in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Challis, C D [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)

    2004-12-01

    Internal transport barriers (ITBs) can provide high tokamak confinement at modest plasma current. This is desirable for operation with most of the current driven non-inductively by the bootstrap mechanism, as currently envisaged for steady-state power plants. Maintaining such plasmas in steady conditions with high plasma purity is challenging, however, due to MHD instabilities and impurity transport effects. Significant progress has been made in the control of ITB plasmas: the pressure profile has been varied using the barrier location; q-profile modification has been achieved with non-inductive current drive, and means have been found to affect density peaking and impurity accumulation. All these features are, to some extent, interdependent and must be integrated self-consistently to demonstrate a sound basis for extrapolation to future devices.

  13. Electron thermal transport barrier and magnetohydrodynamic activity observed in Tokamak plasmas with negative central shear

    NARCIS (Netherlands)

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

    1997-01-01

    In the Rijnhuizen Tokamak Project, plasmas with steady-state negative central shear (NCS) are made with off-axis electron cyclotron heating. Shifting the power deposition by 2 mm results in a sharp transition of confinement. The good confinement branch features a transport barrier at the off-axis

  14. Plasma position control in a tokamak reactor around ignition

    International Nuclear Information System (INIS)

    Carretta, U.; Minardi, E.; Bacelli, N.

    1986-01-01

    Plasma position control in a tokamak reactor in the phase approaching ignition is closely related to burn control. If ignited burn corresponds to a thermally unstable situation the plasma becomes sensitive to the thermal instability already in the phase when ignition is approached so that the trajectory in the position-pressure (R,p) space becomes effectively unpredictable. For example, schemes involving closed cycles around ignition can be unstable in the heating-cooling phases, and the deviations may be cumulative in time. Reliable plasma control in pressure-position (p, R) space is achieved by beforehand constraining the p, R trajectory rigidly with suitable feedback vertical field stabilization, which is to be established already below ignition. A scheme in which ignition is approached in a stable and automatic way by feedback stabilization on the vertical field is proposed and studied in detail. The values of the gain coefficient ensuring stabilization and the associated p and R excursions are discussed both analytically, with a 0-D approximation including non-linear effects, and numerically with a 1-D code in cylindrical geometry. Profile effects increase the excursions, in particular above ignition. (author)

  15. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  16. Initial plasma production by induction electric field on QUEST tokamak

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke

    2007-01-01

    Induction electric field by center solenoid coil plays a roll to produce initial plasma. According to Townsend avalanche theory, minimum electric field for plasma breakdown depends on neutral gas pressure and connection length. On QUEST spherical tokamak, a connection length is evaluated as 966m on null point neighborhood with coil current ratio I PF26 /I CS =0.1, and induction electric field considering eddy current of vacuum vessel is evaluated as about 0.1 V/m on null point neighborhood. With Townsend avalanche theory, these values manage to produce initial plasma on QUEST. (author)

  17. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    International Nuclear Information System (INIS)

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs

  18. Three-dimensional particle-in-cell simulations of gap crossings in castellated plasma-facing components in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Komm, Michael; Dejarnac, Renaud; Gunn, J. P.; Pekarek, Z.

    2013-01-01

    Roč. 55, č. 2 (2013), 025006-025006 ISSN 0741-3335 R&D Projects: GA ČR GA202/09/1467; GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * PIC divertor * castellation gaps Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.386, year: 2013 http://iopscience.iop.org/0741-3335/55/2/025006/pdf/0741-3335_55_2_025006.pdf

  19. Plasma formation and first OH experiments in GLOBUS-M tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Aleksandrov, S.V.; Burtseva, T.A.

    2001-01-01

    The paper reports results of experimental campaigns on plasma ohmic heating, performed during 1999-2000 on the spherical tokamak Globus-M. Later experimental results with tokamak fed by thyristor rectifiers are presented in detail. The toroidal magnetic field and plasma pulse duration in these experiments were significantly increased. The method of stray magnetic field compensation is described. The technology of vacuum vessel conditioning, including boronization of the vessel performed at the end of the experiments, is briefly discussed. Also discussed is the influence of ECR preioniziation on the breakdown conditions. Experimental data on plasma column formation and current ramp-up in different regimes of operation with the magnetic flux of the central solenoid (CS) limited to ∼100 mVs are presented. Ramp-up of the plasma current of 0.25 MA for the time interval ∼0.03 s with about 0.02 s flat-top at the toroidal field (TF) strength of 0.35 T allows the conclusion that power supplies, control system and wall conditioning work well. The same conclusion can be drawn from observation of plasma density behavior the density is completely controlled with external gas puff and the influence of the wall is negligible after boronization. The magnetic flux consumption efficiency is discussed. The results of magnetic equilibrium simulations are presented and compared with experiment. (author)

  20. A role of neutral hydrogen in CHS plasmas with reheat and collapse and comparison with JIPP T-IIU tokamak plasmas

    International Nuclear Information System (INIS)

    Morita, S.; Yamada, H.; Iguchi, H.

    1992-09-01

    Results are described on NBI plasmas of the Compact Helical System (CHS). An increase in the stored energy, which is called plasma 'reheat', is observed with density peaking when gas puffing is turned off in the high density region. A plasma collapse with large increase in radiation loss occurs even in discharges whose Z eff values (typically, less than 2-3) do not show any increase when the gas puffing is continued. Both phenomena are basically explained by the edge electron temperature due to the difference in the amount of edge hydrogen neutrals. After turning off the gas puffing, the central electron density n e0 shows an increase of 80% and the density peaking factor (n e0 /n-bar e ) changes from 1.0 to 2.0, in typical cases, and a high inward velocity of the impurities appears (v = 20 m/s). The accumulation is studied in relation to the poloidal rotation and the edge temperature. These results are compared with results from plasmas with IOC- and H-modes in the JIPPT-IIU tokamak. (author)