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)
Synchronous oscillation prior to disruption caused by kink modes in HL-2A tokamak plasmas
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.
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)
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)
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)
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)
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
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
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
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)
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)
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)
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
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)
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)
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
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)
Control system of neoclassical tearing modes in real time on HL-2A tokamak
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.
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)
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)
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.
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.
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)
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
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.
Optics of ion beams for the neutral beam injection system on HL-2A Tokamak.
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.
Design and construction of high-frequency magnetic probe system on the HL-2A tokamak
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.
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)
Experimental and modeling researches of dust particles in the HL-2A tokamak
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).
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
A new high sensitivity far-infrared laser interferometer for the HL-2A tokamak
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.
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
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)
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)
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)
Research on the phase adjustment method for dispersion interferometer on HL-2A tokamak
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.
Control of edge localized modes by pedestal deposited impurity in the HL-2A tokamak
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.
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)
Characterization of plasma current quench during disruptions at HL-2A
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.
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)
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)
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)
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
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.
Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak
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.
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
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
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.
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)
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)
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
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
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)
The Supervisory Control System for the HL-2A Neutral Beam Injector
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.
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)
Bayesian soft x-ray tomography and MHD mode analysis on HL-2A
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.
ECRH Studies on Tokamak Plasmas.
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
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.
Synergetic Effects of Runaway and Disruption Induced by VDE on the First Wall Damage in HL-2A
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.
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)
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)
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.)
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)
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
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.)
Energy Technology Data Exchange (ETDEWEB)
White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.
1989-01-01
The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.
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
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
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
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.
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
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)
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)
Simulation models for tokamak plasmas
International Nuclear Information System (INIS)
Dimits, A.M.; Cohen, B.I.
1992-01-01
Two developments in the nonlinear simulation of tokamak plasmas are described: (A) Simulation algorithms that use quasiballooning coordinates have been implemented in a 3D fluid code and a 3D partially linearized (Δf) particle code. In quasiballooning coordinates, one of the coordinate directions is closely aligned with that of the magnetic field, allowing both optimal use of the grid resolution for structures highly elongated along the magnetic field as well as implementation of the correct periodicity conditions with no discontinuities in the toroidal direction. (B) Progress on the implementation of a likeparticle collision operator suitable for use in partially linearized particle codes is reported. The binary collision approach is shown to be unusable for this purpose. The algorithm under development is a complete version of the test-particle plus source-field approach that was suggested and partially implemented by Xu and Rosenbluth
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.
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
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
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
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
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
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
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)
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
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.)
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).
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
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
Advanced tokamak burning plasma experiment
International Nuclear Information System (INIS)
Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.
2001-01-01
A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)
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 ...
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
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.
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)
Pumping and leak detection system of the HL-2A
International Nuclear Information System (INIS)
Cao Zeng; Xu Yunxian; Fu Weidong
2001-01-01
The pumping system is a combination of 8 turbomolecular pumps with three stages pumping for HL-2A vacuum vessel, a total effective pumping speed at the vessel of 12 m 3 ·s -1 for nitrogen. The leak detection of element and vessel is performed with inspiration, case of leak detection and two mass spectrometry. The total leak rate of vessel is bellow 1 x 10 -5 Pa ·m 3 ·s -1 . The base pressure is 1 x 10 -5 Pa
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
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 ...
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)
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)
Scrape-off layer tokamak plasma turbulence
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.
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
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
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
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
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.
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
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)
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
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
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
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)
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 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%.
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%
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)
Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas
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
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)
Viscosity in the edge of tokamak plasmas
International Nuclear Information System (INIS)
Stacey, W.M.
1993-05-01
A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the ''short-radial-gradient-scale-length'' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates
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)
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)
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
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
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
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
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
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)
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.
Integrated plasma control for high performance tokamaks
International Nuclear Information System (INIS)
Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.
2005-01-01
Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)
Multiscale coherent structures in tokamak plasma turbulence
International Nuclear Information System (INIS)
Xu, G. S.; Wan, B. N.; Zhang, W.; Yang, Q. W.; Wang, L.; Wen, Y. Z.
2006-01-01
A 12-tip poloidal probe array is used on the HT-7 superconducting tokamak [Li, Wan, and Mao, Plasma Phys. Controlled Fusion 42, 135 (2000)] to measure plasma turbulence in the edge region. Some statistical analysis techniques are used to characterize the turbulence structures. It is found that the plasma turbulence is composed of multiscale coherent structures, i.e., turbulent eddies and there is self-similarity in a relative short scale range. The presence of the self-similarity is found due to the structural similarity of these eddies between different scales. These turbulent eddies constitute the basic convection cells, so the self-similar range is just the dominant scale range relevant to transport. The experimental results also indicate that the plasma turbulence is dominated by low-frequency and long-wavelength fluctuation components and its dispersion relation shows typical electron-drift-wave characteristics. Some large-scale coherent structures intermittently burst out and exhibit a very long poloidal extent, even longer than 6 cm. It is found that these large-scale coherent structures are mainly contributed by the low-frequency and long-wavelength fluctuating components and their presence is responsible for the observations of long-range correlations, i.e., the correlation in the scale range much longer than the turbulence decorrelation scale. These experimental observations suggest that the coexistence of multiscale coherent structures results in the self-similar turbulent state
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)
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
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)
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)
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)
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
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.
The COMPASS Tokamak Plasma Control Software Performance
Valcarcel, Daniel F.; Neto, André; Carvalho, Ivo S.; Carvalho, Bernardo B.; Fernandes, Horácio; Sousa, Jorge; Janky, Filip; Havlicek, Josef; Beno, Radek; Horacek, Jan; Hron, Martin; Panek, Radomir
2011-08-01
The COMPASS tokamak has began operation at the IPP Prague in December 2008. A new control system has been built using an ATCA-based real-time system developed at IST Lisbon. The control software is implemented on top of the MARTe real-time framework attaining control cycles as short as 50 μs, with a jitter of less than 1 μs. The controlled parameters, important for the plasma performance, are the plasma current, position of the plasma current center, boundary shape and horizontal and vertical velocities. These are divided in two control cycles: slow at 500 μs and fast at 50 μs. The project has two phases. First, the software implements a digital controller, similar to the analog one used during the COMPASS-D operation in Culham. In the slow cycle, the plasma current and position are measured and controlled with PID and feedforward controllers, respectively, the shaping magnetic field is preprogrammed. The vertical instability and horizontal equilibrium are controlled with the faster 50-μs cycle PID controllers. The second phase will implement a plasma-shape reconstruction algorithm and controller, aiming at optimized plasma performance. The system was designed to be as modular as possible by breaking the functional requirements of the control system into several independent and specialized modules. This splitting enabled tuning the execution of each system part and to use the modules in a variety of applications with different time constraints. This paper presents the design and overall performance of the COMPASS control software.
Tokamak nonmaxwellian plasma dynamics in thermonuclear regime
International Nuclear Information System (INIS)
Cotsaftis, M.
1987-01-01
To reach ignition in a Tokamak plasma, large additional power P aux has to be injected in the device on top of the Joule heating P OH =VI r , V the plasma loop voltage, I r the resistive port of plasma current. Typi-cally JH ∼ 1 KeV, whereas ignition would requi- re IG ∼ 7-10 KeV. To gain this factor 7, one at least should inject additional power P aux ∼ 7P OH , supposing that nothing, especially the heat transport, is modified. This is by far not the case, with the so-called energy lifetime degradation, largely observed in oil experiments (but less dramatic with divertors), where energy lifetime tau E behaves like P tot -b with b∼1/2. In large machines where ignition temperature is the target to be imperiously reached, this implies to inject a very large power, typically P aux ∼ 50 to 100 MW, depending on size and parameters and on actual transport. So it is of importance with such figures, or even larger ones owing to uncertain ties, to optimize at best injected power by increasing its efficiency, both with respect to possible transport laws, and to physical phenomena governing heat flow in the system from the sources. This leads to the concept of scenarios, as time sequences of power input, where physical properties of the plasma system are used to build up ion temperature so that ignition is reached with minimum P tot = P OH + P aux and with fixed Q = Q o > 1. Elements for this study are given. The method is outlined. The resulting system of equations describing the evolution of a thermonuclear plasma is given
Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA
Indian Academy of Sciences (India)
The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is ﬁrst tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...
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
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].
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)
Plasma Confinement in the UCLA Electric Tokamak.
Taylor, Robert J.
2001-10-01
The main goal of the newly constructed large Electric Tokamak (R = 5 m, a = 1 m, BT 8 x 10^12 cm-3 when there is no MHD activity. The electron temperature, derived from the plasma conductivity is > 250 eV with a central electron energy confinement time > 350 msec in ohmic conditions. The sawteeth period is 50 msec. Edge plasma rotation is induced by plasma biasing via electron injection in an analogous manner to that seen in CCT(R.J. Taylor, M.L. Brown, B.D. Fried, H. Grote, J.R. Liberati, G.J. Morales, P. Pribyl, D. Darrow, and M. Ono. Phys. Rev Lett. 63 2365 1989.) and the neoclassical bifurcation is close to that described by Shaing et al(K.C. Shaing and E.C. Crume, Phys. Rev. Lett. 63 2369 (1989).). In the ohmic phase the confinement tends to be MHD limited. The ICRF heating eliminates the MHD disturbances. Under second harmonic heating conditions, we observe an internal confinement peaking characterized by doubling of the core density and a corresponding increase in the central electron temperature. Charge exchange data, Doppler data in visible H-alpha light, and EC radiation all indicate that ICRF heating works much better than expected. The major effort is focused on increasing the power input and controlling the resulting equilibrium. This task appears to be easy since our current pulses are approaching the 3 second mark without RF heating or current drive. Our initial experience with current profile control, needed for high beta plasma equilibrium, will be also discussed.
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
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
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)
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
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
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)
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
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
Lower hybrid current drive in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1999-03-01
Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar{sub e} - 10{sup 20}m{sup -3}, ALCATOR-C) and the highest current drive efficiency ({eta}{sub CD} = 3.5x10{sup 19} m{sup -2}A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)
Lower hybrid current drive in tokamak plasmas
International Nuclear Information System (INIS)
Ushigusa, Kenkichi
1999-03-01
Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar e - 10 20 m -3 , ALCATOR-C) and the highest current drive efficiency (η CD = 3.5x10 19 m -2 A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)
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.
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)
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
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.
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.)
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.
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
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 ...
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
Turbulent ion heating in TCV Tokamak plasmas
International Nuclear Information System (INIS)
Schlatter, Ch.
2009-08-01
The Tokamak à configuration variable (TCV) features the highest electron cyclotron wave power density available to resonantly heat (ECRH) the electrons and to drive noninductive currents in a fusion grade plasma (ECCD). In more than 15 years of exploitation, much effort has been expended on real and velocity space engineering of the plasma electron energy distribution function and thus making electron physics a major research contribution of TCV. When a plasma was first subjected to ECCD, a surprising energisation of the ions, perpendicular to the confining magnetic field, was observed on the charge exchange spectrum measured with the vertical neutral particle analyser (VNPA). It was soon concluded that the ion acceleration was not due to power equipartition between electrons and ions, which, due to the absence of direct ion heating on TCV, has thus far been considered as the only mechanism heating the ions. However, although observed for more than ten years, little attention was paid to this phenomenon, whose cause has remained unexplained to date. The key subject of this thesis is the experimental study of this anomalous ion acceleration, the characterisation in terms of relevant parameters and the presentation of a model simulation of the potential process responsible for the appearance of fast ions. The installation of a new compact neutral particle analyser (CNPA) with an extended high energy range (≥ 50 keV) greatly improved the fast ion properties diagnosis. The CNPA was commissioned and the information derived from its measurement (ion temperature and density, isotopic plasma composition) was validated against other ion diagnostics, namely the active carbon charge exchange recombination spectroscopy system (CXRS) and a neutron counter. In ohmic plasmas, where the ion heating agrees with classical theory, the radial ion temperature profile was successfully reconstructed by vertically displacing the plasma across the horizontal CNPA line of sight. Active
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
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
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
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.
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
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
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
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
The calculation and simulation of the ECRH HV power supply for the HL-2A tokamak
International Nuclear Information System (INIS)
Mao Xiaohui; Li Qing; Xuan Weimin; Yao Lieying
2006-01-01
In order to satisfy the requirement of ECRH, the ECRH HV power supply (ECHV-HVPS) on the base of high voltage pulse modulation has been designed. The filter inductance in the ECHV-HVPS is much smaller than the voltage regulation power supply. Modulations are adopted in the power supply, so the short time of the leading edge and lagging edge of the pulse is achieved. The main circuit of the ECHV-HVPS is showed. The equivalent resistance and the transient response of the PS are calculated and analyzed using MATLAB, and experiment results are given. (authors)
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
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
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)
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.
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
Experimental methods to study tokamak plasma stability
International Nuclear Information System (INIS)
Perez-Navarro, A.
1978-01-01
Experimental devices to measure external instability modes with small pick-up coils to detect poloidal magnetic field fluctuations, and internal modes with soft-X-ray detectors are discussed. The characteristics of these devices are calculated for a small tokamak (R 0 = 30 cm, a = 10 cm, I 0 50 KA). (author)
[High beta tokamak research and plasma theory
International Nuclear Information System (INIS)
1990-01-01
Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report
Anomalous periodic disruptions in tokamak plasma
International Nuclear Information System (INIS)
Montvai, A.; Tegze, M.; Valyi, I.
1982-09-01
Anomalously strong, periodic instabilities were observed in the MT-1 tokamak. Characteristics of these instabilities were partly similar to those of internal disruptions, but there were features making them different from the normal relaxational oscillations. Basic characteristics of the phenomenon were studied with the aid of generally used diagnostics. (author)
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.
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.
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)
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)
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.
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
Hybrid model for simulation of plasma jet injection in tokamak
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.
Design for the excitation systems of three generator sets of HL-2A
International Nuclear Information System (INIS)
Li Huajun; Xu Lirong; Liu Xuemei; You Tianxue
2001-01-01
The design for the excitation systems have been made on the basis of respective features and load demands of three motor-generator sets of HL-2A. The excitation systems of No.1 and No.2 generator sets which supply for toroidal field are discussed in detail and three feasible excitation plans have been proposed according to the investment
Numerical simulation of MHD equilibrium configuration for the HL-2A modification
International Nuclear Information System (INIS)
Chen Qian; Wang Aike; Li Fangzhu; Zhang Jinghua
2008-01-01
Numerical simulation is employed for the HL-2A modification, which includes the optimum design of zero-field in the start-up phase, the limiter equilibrium configuration, the single/double null divertor equilibrium configuration, and the equilibrium configuration evolution from gas breakdown to current plateau. Results show that the new program can satisfy the design requirement. (authors)
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
Neural net prediction of tokamak plasma disruptions
International Nuclear Information System (INIS)
Hernandez, J.V.; Lin, Z.; Horton, W.; McCool, S.C.
1994-10-01
The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system
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)
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)
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)
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.
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
Plasma shaping effects on tokamak scrape-off layer turbulence
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/{{\
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
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
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.
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.
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
'Snowflake' H Mode in a Tokamak Plasma
International Nuclear Information System (INIS)
Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.
2010-01-01
An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a 'snowflake' (second-order null) divertor configuration for the first time in the TCV tokamak. This regime exhibits 2 to 3 times lower ELM frequency and 20%-30% increased 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. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.
Plasma radiation in tokamak disruption simulation experiments
International Nuclear Information System (INIS)
Arkhipov, N.; Bakhtin, V.; Safronov, V.; Toporkov, D.; Vasenin, S.; Zhitlukhin, A.; Wuerz, H.
1995-01-01
Plasma impact results in sudden evaporation of divertor plate material and produces a plasma cloud which acts as a protective shield. The incoming energy flux is absorbed in the plasma shield and is converted mainly into radiation. Thus the radiative characteristics of the target plasma determine the dissipation of the incoming energy and the heat load at the target. Radiation of target plasma is studied at the two plasma gun facility 2MK-200 at Troitsk. Space- and time-resolved spectroscopy and time-integrated space-resolved calorimetry are employed as diagnostics. Graphite and tungsten samples are exposed to deuterium plasma streams. It is found that the radiative characteristics depend strongly on the target material. Tungsten plasma arises within 1 micros close to the surface and shows continuum radiation only. Expansion of tungsten plasma is restricted. For a graphite target the plasma shield is a mixture of carbon and deuterium. It expands along the magnetic field lines with a velocity of v = (3--4) 10 6 cm/s. The plasma shield is a two zone plasma with a hot low dense corona and a cold dense layer close to the target. The plasma corona emits intense soft x-ray (SXR) line radiation in the frequency range from 300--380 eV mainly from CV ions. It acts as effective dissipation system and converts volumetrically the incoming energy flux into SXR radiation
Sawtooth driven particle transport in tokamak plasmas
International Nuclear Information System (INIS)
Nicolas, T.
2013-01-01
The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr
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)
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)
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.
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
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
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)
MHD Effects of a Ferritic Wall on Tokamak Plasmas
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
Anisotropic plasma with flows in tokamak: Steady state and stability
International Nuclear Information System (INIS)
Ilgisonis, V.I.
1996-01-01
An adequate description of equilibrium and stability of anisotropic plasma with macroscopic flows in tokamaks is presented. The Chew-Goldberger-Low (CGL) approximation is consistently used to analyze anisotropic plasma dynamics. The admissible structure of a stationary flow is found to be the same as in the ideal magnetohydrodynamics with isotropic pressure (MHD), which means an allowance for the same relabeling symmetry as in ideal MHD systems with toroidally nested magnetic surfaces. A generalization of the Grad-Shafranov equation for the case of anisotropic plasma with flows confined in the axisymmetric magnetic field is derived. A variational principle was obtained, which allows for a stability analysis of anisotropic pressure plasma with flows, and takes into account the conservation laws resulting from the relabeling symmetry. This principle covers the previous stability criteria for static CGL plasma and for ideal MHD flows in isotropic plasma as well. copyright 1996 American Institute of Physics
MTX [Microwave Tokamak Experiment] plasma diagnostic system
International Nuclear Information System (INIS)
Rice, B.W.; Hooper, E.B.; Brooksby, C.A.
1987-01-01
In this paper, a general overview of the MTX plasma diagnostics system is given. This includes a description of the MTX machine configuration and the overall facility layout. The data acquisition system and techniques for diagnostic signal transmission are also discussed. In addition, the diagnostic instruments planned for both an initial ohmic-heating set and a second FEL-heating set are described. The expected range of plasma parameters along with the planned plasma measurements will be reviewed. 7 refs., 5 figs
Preliminary application of maximum likelihood method in HL-2A Thomson scattering system
International Nuclear Information System (INIS)
Yao Ke; Huang Yuan; Feng Zhen; Liu Chunhua; Li Enping; Nie Lin
2010-01-01
Maximum likelihood method to process the data of HL-2A Thomson scattering system is presented. Using mathematical statistics, this method maximizes the possibility of the likeness between the theoretical data and the observed data, so that we could get more accurate result. It has been proved to be applicable in comparison with that of the ratios method, and some of the drawbacks in ratios method do not exist in this new one. (authors)
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.)
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
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
International Nuclear Information System (INIS)
Castracane, J.
2001-01-01
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies
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
Momentum Injection in Tokamak Plasmas and Transitions to Reduced Transport
International Nuclear Information System (INIS)
Parra, F. I.; Highcock, E. G.; Schekochihin, A. A.; Barnes, M.; Cowley, S. C.
2011-01-01
The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.
Multi-field plasma sandpile model in tokamaks and applications
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.
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
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Castracane, J.
2001-01-04
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.
Burn stability of tokamak fusion plasmas with synergetic current drive
International Nuclear Information System (INIS)
Anderson, D.; Lisak, M.; Kolesnichenko, Ya.
1991-01-01
The stability of thermonuclear burn in Tokamak-reactors with non-inductive current generated with the simultaneous application of various methods is investigated. Particular emphasis is given to the ITER synergetic current drive scenario involving LH waves, neoclassical effects and NB injection. For ITER-like confinement laws, it is shown that this scenario may be unstable on the plasma skin time scale. Figs
Tokamak plasma current disruption infrared control system
International Nuclear Information System (INIS)
Kugel, H.W.; Ulrickson, M.
1987-01-01
This patent describes a device for magnetically confining a plasma driven by a plasma current and contained within a toroidal vacuum chamber, the device having an inner toroidal limiter on an inside wall of the vacuum chamber and an arrangement for the rapid prediction and control in real time of a major plasma disruption. The arrangement is described which includes: scanning means sensitive to infrared radiation emanating from within the vacuum chamber, the infrared radiation indicating the temperature along a vertical profile of the inner toroidal limiter. The scanning means is arranged to observe the infrared radiation and to produce in response thereto an electrical scanning output signal representative of a time scan of temperature along the vertical profile; detection means for analyzing the scanning output signal to detect a first peaked temperature excursion occurring along the profile of the inner toroidal limiter, and to produce a detection output signal in repsonse thereto, the detection output signal indicating a real time prediction of a subsequent major plasma disruption; and plasma current reduction means for reducing the plasma current driving the plasma, in response to the detection output signal and in anticipation of a subsequent major plasma disruption
Plasma diagnostics for tokamaks and stellarators
Energy Technology Data Exchange (ETDEWEB)
Stott, P E; Sanchez, J
1994-07-01
A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: Magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Scattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma. (Author) 451 refs.
Plasma diagnostics for tokamaks and stellarators
International Nuclear Information System (INIS)
Stott, P.E.; Sanchez, J.
1994-01-01
A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Sattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma
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.
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)
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
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
Liquid gallium jet-plasma interaction studies in ISTTOK tokamak
International Nuclear Information System (INIS)
Gomes, R.B.; Fernandes, H.; Silva, C.; Sarakovskis, A.; Pereira, T.; Figueiredo, J.; Carvalho, B.; Soares, A.; Duarte, P.; Varandas, C.; Lielausis, O.; Klyukin, A.; Platacis, E.; Tale, I.; Alekseyv, A.
2009-01-01
Liquid metals have been pointed out as a suitable solution to solve problems related to the use of solid walls submitted to high power loads allowing, simultaneously, an efficient heat exhaustion process from fusion devices. The most promising candidate materials are lithium and gallium. However, lithium has a short liquid state temperature range when compared with gallium. To explore further this property, ISTTOK tokamak is being used to test the interaction of a free flying liquid gallium jet with the plasma. ISTTOK has been successfully operated with this jet without noticeable discharge degradation and no severe effect on the main plasma parameters or a significant plasma contamination by liquid metal. Additionally the response of an infrared sensor, intended to measure the jet surface temperature increase during its interaction with the plasma, has been studied. The jet power extraction capability is extrapolated from the heat flux profiles measured in ISTTOK plasmas.
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)
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
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
Pellet-plasma interactions in tokamaks
DEFF Research Database (Denmark)
Chang, C.T.
1991-01-01
confinement time, offset by the accumulation of impurities at the plasma core is brought into focus. A possible remedy is suggested to diminish the effect of the impurities. Plausible arguments are presented to explain the apparent controversial observations on the propagation of a fast cooling front ahead......The ablation of a refuelling pellet of solid hydrogen isotopes is governed by the plasma state, especially the density and energy distribution of the electrons. On the other hand, the cryogenic pellet gives rise to perturbations of the plasma temperature and density. Based on extensive experimental...... data, the interaction between the pellet and the plasma is reviewed. Among the subjects discussed are the MHD activity, evolution of temperature and density profiles, and the behaviour of impurities following the injection of a pellet (or pellets). The beneficial effect of density peaking on the energy...
Electron cyclotron emission imaging in tokamak plasmas
Munsat, T.; Domier, C.W.; Kong, X. Y.; Liang, T. R.; N C Luhmann Jr.,; Tobias, B. J.; Lee, W.; Park, H. K.; Yun, G.; Classen, I.G.J.; Donne, A. J. H.
2010-01-01
We discuss the recent history and latest developments of the electron cyclotron emission imaging diagnostic technique, wherein electron temperature is measured in magnetically confined plasmas with two-dimensional spatial resolution. The key enabling technologies for this technique are the
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
Evaluation of the plasma parameters in COMPASS tokamak divertor area
Czech Academy of Sciences Publication Activity Database
Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud
2012-01-01
Roč. 356, č. 1 (2012), s. 012007 ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf
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
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
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)
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)
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
Tokamak plasma variations under rapid compression
International Nuclear Information System (INIS)
Holmes, J.A.; Peng, Y.K.M.; Lynch, S.J.
1980-04-01
Changes in plasmas undergoing large, rapid compressions are examined numerically over the following range of aspect ratios A:3 greater than or equal to A greater than or equal to 1.5 for major radius compressions of circular, elliptical, and D-shaped cross sections; and 3 less than or equal to A less than or equal to 6 for minor radius compressions of circular and D-shaped cross sections. The numerical approach combines the computation of fixed boundary MHD equilibria with single-fluid, flux-surface-averaged energy balance, particle balance, and magnetic flux diffusion equations. It is found that the dependences of plasma current I/sub p/ and poloidal beta anti β/sub p/ on the compression ratio C differ significantly in major radius compressions from those proposed by Furth and Yoshikawa. The present interpretation is that compression to small A dramatically increases the plasma current, which lowers anti β/sub p/ and makes the plasma more paramagnetic. Despite large values of toroidal beta anti β/sub T/ (greater than or equal to 30% with q/sub axis/ approx. = 1, q/sub edge/ approx. = 3), this tends to concentrate more toroidal flux near the magnetic axis, which means that a reduced minor radius is required to preserve the continuity of the toroidal flux function F at the plasma edge. Minor radius compressions to large aspect ratio agree well with the Furth-Yoshikawa scaling laws
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)
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
Thermal stability of the tokamak plasma edge
International Nuclear Information System (INIS)
Stacey, W.M.
1997-01-01
The general linear, fluid, thermal instability theory for the plasma edge has been extended. An analysis of a two-dimensional fluid model of the plasma edge has identified the importance of many previously unappreciated phenomena associated with parallel and gyroviscous forces in the presence of large radial gradients, with large radial or parallel flows, with the temperature dependence of transport coefficients, and with the coupling of temperature, flow and density perturbations. The radiative condensation effect is generalized to include a further destabilizing condensation effect associated with radial heat conduction. Representative plasma edge neutral and impurity densities are found to be capable of driving thermal instabilities in the edge transport barrier and radiative mantle, respectively. (author)
Extremely shaped plasmas to improve the Tokamak concept
Energy Technology Data Exchange (ETDEWEB)
Piras, F.
2011-04-15
experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is
Extremely shaped plasmas to improve the Tokamak concept
International Nuclear Information System (INIS)
Piras, F.
2011-04-01
experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is
The COMPASS Tokamak Plasma Control Software Performance
Czech Academy of Sciences Publication Activity Database
Valcárcel, D.F.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Janky, F.; Havlíček, Josef; Beňo, R.; Horáček, Jan; Hron, Martin; Pánek, Radomír
2011-01-01
Roč. 58, č. 4 (2011), s. 1490-1496 ISSN 0018-9499. [Real Time Conference, RT10/17th./. Lisboa, 24.05.2010-28.05.2010] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-Time * ATCA * Data Acquisition * Plasma Control Software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.447, year: 2011 http://dx.doi.org/10.1109/TNS.2011.2143726
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
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.
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)
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
Probabilistic analysis of tokamak plasma disruptions
International Nuclear Information System (INIS)
Sanzo, D.L.; Apostolakis, G.E.
1985-01-01
An approximate analytical solution to the heat conduction equations used in modeling component melting and vaporization resulting from plasma disruptions is presented. This solution is then used to propagate uncertainties in the input data characterizing disruptions, namely, energy density and disruption time, to obtain a probabilistic description of the output variables of interest, material melted and vaporized. (orig.)
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
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
Online Plasma Shape Reconstruction for EAST Tokamak
International Nuclear Information System (INIS)
Luo Zhengping; Xiao Bingjia; Zhu Yingfei; Yang Fei
2010-01-01
An online plasma shape reconstruction, based on the offline version of the EFIT code and MPI library, can be carried out between two adjacent shots in EAST. It combines online data acquisition, parallel calculation, and data storage together. The program on the master node of the cluster detects the termination of the discharge promptly, reads diagnostic data from the EAST mdsplus server on the completion of data storing, and writes the results onto the EFIT mdsplus server after the calculation is finished. These processes run automatically on a nine-nodes IBM blade center. The total time elapsed is about 1 second to several minutes, depending on the duration of the shot. With the results stored in the mdsplus server, it is convenient for operators and physicists to analyze the behavior of plasma using visualization tools.
Plasma shape experiments for an optimized tokamak
International Nuclear Information System (INIS)
Hyatt, A.W.; Osborne, T.H.; Lazarus, E.A.
1994-07-01
In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, βτ E , in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed ''slot'' divertor geometry. power flux to the divertor floor using a closed ''slot'' divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes βτ E and divertor performance. The four shapes studied form a matrix of moderate and high elongations (κ congruent 1.8 and 2.1) and low and high triangularities (δ congruent 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot's transient normalized performance, β N H, where β N ≡ β/(I p )/aB T and H ≡ τ E /τ E ITER-89P , increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D
Plasma shape experiments for an optimized tokamak
Energy Technology Data Exchange (ETDEWEB)
Hyatt, A.W.; Osborne, T.H. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States)
1994-12-31
In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, {beta}{sub {tau}E}, in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed `slot` divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes {beta}{sub {tau}E} and divertor performance. The four shapes studied form a matrix of moderate and high elongations ({kappa} {approx_equal} 1.8 and 2.1) and low and high triangularities ({delta} {approx_equal} 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot`s transient normalized performance, {beta}{sub N}H, where {beta}{sub N} = {beta}/(I{sub p}/aB{sub T}) and H = {tau}{sub E}/{tau}{sub E}{sup ITER-89P}, increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D. (author) 7 refs., 7 figs.
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
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
Two-dimensional transport of tokamak plasmas
International Nuclear Information System (INIS)
Hirshman, S.P.; Jardin, S.C.
1979-01-01
A reduced set of two-fluid transport equations is obtained from the conservation equations describing the time evolution of the differential particle number, entropy, and magnetic fluxes in an axisymmetric toroidal plasma with nested magnetic surfaces. Expanding in the small ratio of perpendicular to parallel mobilities and thermal conductivities yields as solubility constraints one-dimensional equations for the surface-averaged thermodynamic variables and magnetic fluxes. Since Ohm's law E +u x B =R', where R' accounts for any nonideal effects, only determines the particle flow relative to the diffusing magnetic surfaces, it is necessary to solve a single two-dimensional generalized differential equation, (partial/partialt) delpsi. (delp - J x B) =0, to find the absolute velocity of a magnetic surface enclosing a fixed toroidal flux. This equation is linear but nonstandard in that it involves flux surface averages of the unknown velocity. Specification of R' and the cross-field ion and electron heat fluxes provides a closed system of equations. A time-dependent coordinate transformation is used to describe the diffusion of plasma quantities through magnetic surfaces of changing shape
Heavy Neutral Beam Probe for edge plasma analysis in tokamaks
International Nuclear Information System (INIS)
1991-01-01
The Heavy Neutral Beam Probe project presented in this document is part of an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadian de Fusion Magnetique. The overall objective of the effort is to apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes facility in Montreal, Canada. To achieve this goal, a research and development project was started in December, 1990 to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present, satisfactory progress has been achieved. The ion gun is fully operational with the neutralizer in the final assembly stage in preparation for testing. The beam diagnostics have been completed and mounted in the computer automated test stand. The analyzer design and detailed trajectory calculations are nearing completion to allow for the vacuum interface construction. The CAMAC based data acquisition system hardware was integrated into the test stand. Part of this hardware is a component of the Tokamak de Varennes' contribution to the collaboration. Next steps on the critical path include the beginning of the neutralization tests and the start of the analyzer construction. Anticipated installation of the diagnostic on the tokamak is Spring 1992
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.)
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
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
The evolution of the plasma current during tokamak disruptions
International Nuclear Information System (INIS)
Helander, P.; Andersson, F.; Anderson, D.; Lisak, M.; Eriksson, L.G.
2004-01-01
In a tokamak disruption, the ohmic plasma current is partly replaced by a current carried by runaway electrons. This process is analysed by combining the equations for runaway electron generation with Maxwell's equations for the evolution of the electric field. This allows a quantitative understanding to be gained of runaway production in present experiments, and extrapolation to be made to ITER. The runaway current typically becomes more peaked on the magnetic axis than the pre-disruption current. In fact, the central current density can rise although the total current falls, which may have implications for post-disruption plasma stability. Furthermore, it is found that the runaway current easily spreads radially in a filament way due to the high sensitivity of the runaway generation efficiency to plasma parameters. (authors)
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)
High power RF heating and nonthermal distributions in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Peeters, A.G.
1994-12-13
This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL).
High power RF heating and nonthermal distributions in tokamak plasmas
International Nuclear Information System (INIS)
Peeters, A.G.
1994-01-01
This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL)
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
Electromagnetic effects on trace impurity transport in tokamak plasmas
Hein, T.; Angioni, C.
2010-01-01
The impact of electromagnetic effects on the transport of light and heavy impurities in tokamak plasmas is investigated by means of an extensive set of linear gyrokinetic numerical calculations with the code GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and of analytical derivations with a fluid model. The impurity transport is studied by appropriately separating diffusive and convective contributions, and conditions of background microturbulence dominated by both ion temperature gradient (ITG) and trapped electron modes (TEMs) are analyzed. The dominant contribution from magnetic flutter transport turns out to be of pure convective type. However it remains small, below 10% with respect to the E ×B transport. A significant impact on the impurity transport due to an increase in the plasma normalized pressure parameter β is observed in the case of ITG modes, while for TEM the overall effect remains weak. In realistic conditions of high β plasmas in the high confinement (H-) mode with dominant ITG turbulence, the impurity diffusivity is found to decrease with increasing β in qualitative agreement with recent observations in tokamaks. In contrast, in these conditions, the ratio of the total off-diagonal convective velocity to the diagonal diffusivity is not strongly affected by an increase in β, particularly at low impurity charge, due to a compensation between the different off-diagonal contributions.
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)
Ignition and burn control in tokamak plasmas
International Nuclear Information System (INIS)
Borrass, K.; Gruber, O.; Lackner, K.; Minardi, E.; Neuhauser, J.; Wilhelm, R.; Wunderlich, R.; Bromberg, L.; Cohn, D.R.
1981-01-01
Different schemes for the control of the thermal instability in an ignited fusion reactor are analysed by zero- and one-dimensional models. Passive stabilization methods considered are ripple-enhanced ion heat conduction, the effect of the major-radius variation of the plasma column in a time-independent vertical field, and the combination of both effects, including the spatial variation of the toroidal-ripple amplitude. Active control methods analysed are high-Q-driven operation and feedback-controlled major-radius variation following different scenarios. One-dimensional analyses taking into account only conductive losses show the existence of a single unstable mode in the energy balance, justifying, under these assumptions, the study of only global control. (author)
Study of edge turbulence in tokamak plasmas
International Nuclear Information System (INIS)
Sarazin, Y.
1997-01-01
The aim of this work is to propose a new frame to study turbulent transport in plasmas. In order to avoid the restraint of scale separability the forcing by flux is used. A critical one-dimension self-organized cellular model is developed. In keeping with experience the average transport can be described by means of diffusion and convection terms whereas the local transport could not. The instability due to interchanging process is thoroughly studied and some simplified equations are derived. The proposed model agrees with the following experimental results: the relative fluctuations of density are maximized on the edge, the profile shows an exponential behaviour and the amplitude of density fluctuations depends on ionization source strongly. (A.C.)
Characterizing electrostatic turbulence in tokamak plasmas with high MHD activity
Energy Technology Data Exchange (ETDEWEB)
Guimaraes-Filho, Z O; Santos Lima, G Z dos; Caldas, I L; Nascimento, I C; Kuznetsov, Yu K [Instituto de Fisica, Universidade de Sao Paulo, Caixa Postal 66316, 05315-970, Sao Paulo, SP (Brazil); Viana, R L, E-mail: viana@fisica.ufpr.b [Departamento de Fisica, Universidade Federal do Parana, Caixa Postal 19044, 81531-990, Curitiba, PR (Brazil)
2010-09-01
One of the challenges in obtaining long lasting magnetic confinement of fusion plasmas in tokamaks is to control electrostatic turbulence near the vessel wall. A necessary step towards achieving this goal is to characterize the turbulence level and so as to quantify its effect on the transport of energy and particles of the plasma. In this paper we present experimental results on the characterization of electrostatic turbulence in Tokamak Chauffage Alfven Bresilien (TCABR), operating in the Institute of Physics of University of Sao Paulo, Brazil. In particular, we investigate the effect of certain magnetic field fluctuations, due to magnetohydrodynamical (MHD) instabilities activity, on the spectral properties of electrostatic turbulence at plasma edge. In some TCABR discharges we observe that this MHD activity may increase spontaneously, following changes in the edge safety factor, or after changes in the radial electric field achieved by electrode biasing. During the high MHD activity, the magnetic oscillations and the plasma edge electrostatic turbulence present several common linear spectral features with a noticeable dominant peak in the same frequency. In this article, dynamical analyses were applied to find other alterations on turbulence characteristics due to the MHD activity and turbulence enhancement. A recurrence quantification analysis shows that the turbulence determinism radial profile is substantially changed, becoming more radially uniform, during the high MHD activity. Moreover, the bicoherence spectra of these two kinds of fluctuations are similar and present high bicoherence levels associated with the MHD frequency. In contrast with the bicoherence spectral changes, that are radially localized at the plasma edge, the turbulence recurrence is broadly altered at the plasma edge and the scrape-off layer.
Continuous, saturation, and discontinuous tokamak plasma vertical position control systems
Energy Technology Data Exchange (ETDEWEB)
Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)
2016-10-15
Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.
Continuous, saturation, and discontinuous tokamak plasma vertical position control systems
International Nuclear Information System (INIS)
Mitrishkin, Yuri V.; Pavlova, Evgeniia A.; Kuznetsov, Evgenii A.; Gaydamaka, Kirill I.
2016-01-01
Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.
Magnetic fluctuations in the plasma of KT-5C tokamak
International Nuclear Information System (INIS)
Lu Ronghua; Pan Gesheng; Wang Zhijiang; Wen Yizhi; Yu Changxuan; Wan Shude; Liu Wandong; Wang Jun; Xu Min; Xiao Delong; Yu Yi
2004-01-01
A newly developed moveable magnetic probe array was installed on KT-5C tokamak. The profiles of radial and poloidal magnetic fluctuations of the plasma have been measured for (0.5r/a1.1). The experimental results indicate that there is a radial gradient which is greater than relative electrostatic fluctuations and the magnetic fluctuations contribute a little to losses. A strong coherence between fluctuations of 4 mm nearby two points suggests that the magnetic fluctuations have quite a long correlation length
Electron Heating of LHCD Plasma in HT-7 Tokamak
International Nuclear Information System (INIS)
Ding Yonghua; Wan Baonian; Lin Shiyao; Chen Zhongyong; Hu Xiwei; Shi Yuejiang; Hu Liqun; Kong Wei; Zhang Xiaoqing
2006-01-01
Electron heating via lower hybrid current drive (LHCD) has been investigated in HT-7 superconducting tokamak. Experiments show that the central electron temperature T e0 , the volume averaged electron temperature e > and the peaking factor of the electron temperature Q Te = T e0 / e > increase with the lower hybrid wave (LHW) power. Simultaneously the electron heating efficiency and the electron temperature as the function of the central line-averaged electron density (n e ) and the plasma current (I p ) have also been investigated. The experimental results are in a good agreement with those of the classical collision theory and the LHW power deposition theory
Self-consistent treatment of transport in tokamak plasmas
International Nuclear Information System (INIS)
Wilhelmsson, H.
1993-01-01
A theory is developed for the dynamics of tokamak plasmas considering the influence of combinations of simultaneous heating processes (alpha particle, auxiliary and ohmic), thermal conduction and particle diffusion, thermal and particle pinches, thermalization of alpha particles as well as the effects of boundary conditions. The analysis is based on a generalization of the central expansion technique which transforms the partial differential equations to a set of nonlinear coupled equations in time for the dynamic variables. Oscillatory solutions are found, but only in the presence of alpha particle heating. Examples of extensive computer simulations are included which support and complete the analytic results. (26 refs.)
Direction of Impurity Pinch and Auxiliary Heating in Tokamak Plasmas
International Nuclear Information System (INIS)
Angioni, C.; Peeters, A.G.
2006-01-01
A mechanism of particle pinch for trace impurities in tokamak plasmas, arising from the effect of parallel velocity fluctuations in the presence of a turbulent electrostatic potential, is identified analytically by means of a reduced fluid model and verified numerically with a gyrokinetic code for the first time. The direction of such a pinch reverses as a function of the direction of rotation of the turbulence in agreement with the impurity pinch reversal observed in some experiments when moving from dominant auxiliary ion heating to dominant auxiliary electron heating
Real-Time Software for the Compass Tokamak Plasma Control
Energy Technology Data Exchange (ETDEWEB)
Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Prague (Costa Rica)
2009-07-01
This poster presents the flexible and high-performance real time system that guarantees the desired time cycles for plasma control on the COMPASS tokamak: 500 {mu}s for toroidal field, current, equilibrium and shaping; 50 {mu}s for fast control of the equilibrium and vertical instability. This system was developed on top of a high-performance processor and a software framework (MARTe) tailored for real-time. The preliminary measurements indicate that the time constraints will be met on the final solution. The system allows the making of modifications in the future to improve software components. (A.C.)
A theory of the coherent fundamental plasma emission in Tokamaks
International Nuclear Information System (INIS)
Alves, M.V.; Chian, A.C.-L.
1987-01-01
A theoretical model of coherent radiation near the fundamental plasma frequency in tokamaks is proposed. It is shown that, in the presence of runaway electrons, the beam-generated Langmuir waves (L) can be parametrically converted into electromagnetic waves (T) through ponderomotive coupling to ion acoustic waves (S). Two types of pumps are considered: travelling wave pump and standing wave pump. Expressions are derived for the excitation conditions and the growth rates of electromagnetic decay instabilities (L-> T + S), electromagnetic fusion instabilities (L + S -> T) and electromagnetic oscillating two-stream instabilities (L -> T+- S * , where S * is a purely growing mode). (author) [pt
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
Power supplies for plasma column control in COMPASS tokamak
Czech Academy of Sciences Publication Activity Database
Havlíček, Josef; Hauptmann, R.; Peroutka, Oldřich; Tadros, Momtaz; Hron, Martin; Janky, Filip; Vondráček, Petr; Cahyna, Pavel; Mikulín, Ondřej; Šesták, David; Junek, Pavel; Pánek, Radomír
2013-01-01
Roč. 88, 9-10 (2013), s. 1640-1645 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * Power supplies * Feedback control * Vertical displacement * Vertical kicks Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001543#
A theory of the coherent fundamental plasma emission in Tokamaks
International Nuclear Information System (INIS)
Alves, M.V.; Chian, A.C.-L.
1987-07-01
A theoretical model of coherent radiation near the fundamental plasma frequency in Tokamaks is proposed. It is shown that, in the presence of runaway electrons, the beam-generated Langmuir waves (L) can be paarmetrically converted into electromagnetic waves (T) through ponderomotive coupling to ion acoustic waves (S). Two types of pumps are considered: traveling wave and standing wave pump. Expressions are derived for the excitation conditions and the growth rates of electomagnetic decay instabilities (L → T + S), electromagnetic fusion instabilities (L + S → T) and electromagnetic oscillating two-stream instabilities (L → T+-S sup(*) is a purely growing mode). (author) [pt
Magnetohydrodynamic stability of tokamak plasmas with poloidal mode coupling
International Nuclear Information System (INIS)
Shigueoka, H.; Sakanaka, P.H.
1987-01-01
The stability behavior with respect to internal modes is examined for a class of tokamak equilibria with non-circular cross sections. The surfaces of the constant poloidal magnetic flux ψ (R,Z) are obtained numerically by solving the Grad-Shafranov's equation with a specified shape for the outmost plasma surface. The equation of motion for ideal MHD stability is written in a ortogonal coordinate system (ψ, χ, φ). Th e stability analysis is performance numerically in a truncated set of coupled m (poloidal wave number) equations. The calculations involve no approximations, and so all parameters of the equilibrium solution can be arbitrarily varied. (author) [pt
Plasma engineering analyses of tokamak reactor operating space
International Nuclear Information System (INIS)
Houlberg, W.; Attenberger, S.E.
1981-01-01
A comprehensive method is presented for analyzing the potential physics operating regime of fusion reactor plasmas with detailed transport codes. Application is made to the tokamak Fusion Engineering Device (FED). The relationships between driven and ignited operation and supplementary heating requirements are examined. The reference physics models give a finite range of density and temperature over which physics objectives can be reached. Uncertainties in the confinement scaling and differences in supplementary heating methods can expand or contract this operating regime even to the point of allowing ignition with the more optimistic models
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)
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
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.)
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
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
Effect of neutral atoms on tokamak edge plasmas
International Nuclear Information System (INIS)
Fueloep, T.; Catto, Peter J.; Helander, P.
2001-01-01
Neutral atoms can significantly influence the physics of tokamak edge plasmas, e.g., by affecting the radial electric field and plasma flow there, which may, in turn, be important for plasma confinement. Earlier work [Fueloep et al., Phys. Plasmas 5, 3969 (1998)], assuming short mean-free path neutrals and Pfirsch-Schlueter ions, has shown that the ion-neutral coupling through charge-exchange affects the neoclassical flow velocity significantly. However, the mean-free path of the neutrals is not always small in comparison with the radial scale length of densities and temperatures in the edge pedestal. It is therefore desirable to determine what happens in the limit when the neutral mean-free path is comparable with the scale length. In the present work a self-similar solution for the neutral distribution function allowing for strong temperature and density variation is used, following the analysis of Helander and Krasheninnikov [Phys. Plasmas 3, 226 (1995)]. The self-similar solution is possible if the ratio of the mean-free path to the temperature and density scale length is constant throughout the edge plasma. The resulting neutral distribution function is used to investigate the neutral effects on the ion flow and electrostatic potential as this ratio varies from much less than one to order unity
Plasma confinement using biased electrode in the TCABR tokamak
International Nuclear Information System (INIS)
Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.; Machida, M.
2005-01-01
Experimental data obtained on the TCABR tokamak (R = 0.61 m, r = 0.18 m) with an electrally polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1), and hydrogen gas injection adjusted for keeping the electron density at 1.0x10(19) m(-3) without bias. Temporal and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the density, up to a maximum factor of 2.6, while H-alpha hydrogen spectral line intensity decreases, and the CIII impurity stays on the same level. The analysis of temporal and radial profiles of plasma parameters indicates that the confined plasma entered in the H-mode regime. The data analysis shows a maximum enhanced confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the plasma edge, using measured floating potentials and ion saturation currents, show strong decrease in the power spectra and transport. Bifurcation was not observed, and the decrease in the saturation current occurs in 50 microseconds. (author)
Heavy Neutral Beam Probe for edge plasma analysis in Tokamaks
International Nuclear Information System (INIS)
Castracane, J.; Saravia, E.; Beckstead, J.; Aceto, S.
1993-01-01
The contents of this report present the progress achieved to date on the Heavy Neutral Beam Probe project. This effort is an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadien de Fusion Magnetique (CCFM). The overall objective of the effort is to develop and apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes (TdeV) facility in Montreal, Canada. To achieve this goal, a research and development project was established to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present the project is in the middle of its second budget period with the instrumentation on-site at TdeV. The first half of this budget period was used to complete total system tests at InterScience, Inc., dismantle and ship the hardware to TdeV, re-assemble and install the HNBP on the tokamak. Integration of the diagnostic into the TdeV facility has progressed to the point of first beam production and measurement on the plasma. At this time, the HNBP system is undergoing final de-bugging prior to re-start of machine operation in early Fall of this year
Anomalous energy transport in hot plasmas: solar corona and Tokamak
International Nuclear Information System (INIS)
Beaufume, P.
1992-04-01
Anomalous energy transport is studied in two hot plasmas and appears to be associated with a heating of the solar corona and with a plasma deconfining process in tokamaks. The magnetic structure is shown to play a fundamental role in this phenomenon through small scale instabilities which are modelized by means of a nonlinear dynamical system: the Beasts' Model. Four behavior classes are found for this system, which are automatically classified in the parameter space thanks to a neural network. We use a compilation of experimental results relative to the solar corona to discuss current-based heating processes. We find that a simple Joule effect cannot provide the required heating rates, and therefore propose a dimensional model involving a resistive reconnective instability which leads to an efficient and discontinuous heating mechanism. Results are in good agreement with the observations. We give an analytical expression for a diffusion coefficient in tokamaks when magnetic turbulence is perturbing the topology, which we validate thanks to the standard mapping. A realistic version of the Beasts' Model allows to test a candidate to anomalous transport: the thermal filamentation instability
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.)
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)
Two dimensional neutral transport analysis in tokamak plasma
International Nuclear Information System (INIS)
Shimizu, Katsuhiro; Azumi, Masafumi
1987-02-01
Neutral particle influences the particle and energy balance, and play an important role on sputtering impurity and the charge exchange loss of neutral beam injection. In order to study neutral particle behaviour including the effects of asymmetric source and divertor configuration, the two dimensional neutral transport code has been developed using the Monte-Carlo techniques. This code includes the calculation of the H α radiation intensity based on the collisional-radiation model. The particle confinement time of the joule heated plasma in JT-60 tokamak is evaluated by comparing the calculated H α radiation intensity with the experimental data. The effect of the equilibrium on the neutral density profile in high-β plasma is also investigated. (author)
Remote network control plasma diagnostic system for Tokamak T-10
International Nuclear Information System (INIS)
Troynov, V I; Zimin, A M; Krupin, V A; Notkin, G E; Nurgaliev, M R
2016-01-01
The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet. (paper)
Characterisation of detached plasmas on the MAST tokamak
Energy Technology Data Exchange (ETDEWEB)
Harrison, J.R., E-mail: james.harrison@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Lisgo, S.W. [ITER Organization, Route de Vinon-sur-Verdon, St.Paul-lez-Durance, Cedex (France); Gibson, K.J. [Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Dowling, J. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)
2011-08-01
Divertor detachment is an attractive operating regime for the next generation of tokamak devices, as it offers a means of mitigating the steady-state heat flux to plasma facing components. In order to clarify the dominant physical mechanisms that govern detachment, high quality data from several diagnostics are required to constrain theoretical models. To that end, high spatial ({approx}3 mm) and temporal (5 kHz) resolution measurements have been made of the intensity of deuterium Balmer and carbon emission lines during the onset and evolution of detachment of the lower inner strike point in MAST L-mode discharges. Furthermore, spatially-resolved measurements of the shapes and intensities of high-n Balmer lines have been recorded to infer plasma conditions during the detached phase.
Heat flow during sawtooth collapse in tokamak plasmas
International Nuclear Information System (INIS)
Hanada, Kazuaki
1994-01-01
Heat flow during sawtooth collapse was studied on the WT-3 tokamak by using temporal evolution of soft X-ray intensity profile in the poloidal cross section in a lower hybrid current driven plasma as well as an electron cyclotron heated plasma. Two phase in sawtooth collapses were observed. In the first phases, the hottest spot that is the peak of the soft X-ray distribution approaches the inversion surface and heat flows out through a narrow gate on the inversion surface. In the second phase, the hottest spot stays on the inversion surface, and heat flows out through the whole inversion surface. This suggests that magnetic reconnection as predicted by Kadomtsev's model occurs in the first phase, but in the second phase, a different mechanism dominates heat flow. (author)
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.
LH-power coupling in advanced tokamak plasmas in JET
International Nuclear Information System (INIS)
Joffrin, E.; Erents, K.; Gormezano, C.
2000-02-01
Lower Hybrid Current Drive (LHCD) is the most efficient tool to generate non-inductive current in tokamak plasmas. In JET, significant modifications of the current profile have been recently achieved in coupling up to 3MW of LH power in optimised shear discharges. However, the improved particle confinement during optimised shear plasmas results in a sharp decrease of the electron density in front the launcher close or below the cut-off density (ne=1.7.10 17 m -3 for f LH =37GHz) and makes difficult the coupling of the LH power. Deuterium gas near the launcher can help to improve the coupling, but has also the effect of increasing the ELM activity leading to the erosion of the internal transport barrier (ITB). Future development of lower hybrid launcher should include the constraints imposed by scenario such as the optimised shear. (author)
Study of the plasma edge turbulence in tokamaks
International Nuclear Information System (INIS)
Garbet, X.; Laurent, L.; Mourgues, F.; Roubin, J.P.; Samain, A.
1990-01-01
The plasma edge in tokamaks is known to be very turbulent. We investigate here the non linear stability of a test mode in presence of an helical potential perturbation, i.e. a pump mode, which simulates the plasma turbulence. The particle trajectories in this perturbed equilibrium are derived using an hamiltonian formalism. The electrons appear to have trapped trajectories in the potential well of the pump mode, while the ions experience a large convective motion. These two effects have a large influence on the test mode stability. First, non linearly trapped electrons supply an energy source for the test mode. Second, the ion convective motion introduces a radial scale of the test mode larger than the ion Larmor radius, in agreement with experimental data. These two phenomena allow a bifurcation in the turbulence level and provide therefore an explanation for the L-H transition
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.)
Plasma confinement using biased electrode in the TCABR tokamak
International Nuclear Information System (INIS)
Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Machida, M.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.
2005-01-01
Experimental data obtained on the TCABR tokamak (R = 0.61 m, a = 0.18 m) with an electrically polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1) and hydrogen gas injection adjusted for keeping the electron density at 1.0 x 10 19 m -3 without bias. Time evolution and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the central line-averaged density, up to a maximum factor of 2.6, while H α hydrogen spectral line intensity decreases and the C III impurity stays on the same level. The analysis of temporal behaviour and radial profiles of plasma parameters indicates that the confined plasma enters the H-mode regime. The data analysis shows a maximum enhanced energy confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the Scrape-Off-Layer, using measured floating potentials and ion saturation currents, show a strong decrease in the power spectra and transport. Bifurcation was not observed and the decrease in the saturation current occurs in 50 μs
Edge Plasma Response to Non-Axisymmetric Fields in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Ferraro, N. M.; Lao, L. L.; Buttery, R. J.; Evans, T. E.; Snyder, P. B.; Wade, M.R., E-mail: ferraro@fusion.gat.com [General Atomics, San Diego (United States); Moyer, R. A.; Orlov, D. M. [University of California San Diego, La Jolla (United States); Lanctot, M. J. [Lawrence Livermore National Laboratory, Livermore (United States)
2012-09-15
Full text: The application of non-axisymmetric fields is found to have significant effects on the transport and stability of H-mode tokamak plasmas. These effects include dramatic changes in rotation and particle transport, and may lead to the partial or complete suppression of edge-localized modes (ELMs) under some circumstances. The physical mechanism underlying these effects is presently not well understood, in large part because the response of the plasma to non- axisymmetric fields is significant and complex. Here, recent advances in modeling the plasma response to non-axisymmetric fields are discussed. Calculations using a resistive two-fluid model in diverted toroidal geometry confirm the special role of the perpendicular electron velocity in suppressing the formation of islands in the plasma. The possibility that islands form near the top of the pedestal, where the zero-crossing of the perpendicular electron velocity may coincide with a mode-rational surface, is explored, and the implications for ELM suppression are discussed. Modeling results are compared with empirical data. It is shown that numerical modeling is successful in reproducing some experimentally observed effects of applied non-axisymmetric fields on the edge temperature and density profiles. The numerical model self-consistently includes the plasma, separatrix, and scrape-off layer. Rotation and diamagnetic effects are also included self-consistently. Solutions are calculated using the M3D-C1 extended-MHD code. (and others)
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
Analysis of equilibrium and topology of tokamak plasmas
International Nuclear Information System (INIS)
Milligen, B.P. van.
1991-01-01
In a tokamak, the plasma is confined by means of a magnetic field. There exists an equilibrium between outward forces due to the pressure gradient in plasma and inward forces due to the interaction between currents flowing inside the plasma and the magnetic field. The equilibrium magnetic field is characterized by helical field lines that lie on nested toroidal surfaces of constant flux. The equilibrium yields values for global and local plasma parameters (e.g. plasma position, total current, local pressure). Thus, precise knowledge of the equilibrium is essential for plasma control, for the understanding of many phenomena occurring in the plasma (in particular departures from the ideal equilibrium involving current filamentation on the flux surfaces that lead to the formation of islands, i.e. nested helical flux surfaces), and for the interpretation of many different types of measurements (e.g. the translation of line integrated electron density measurements made by laser beams probing the plasma into a local electron density on a flux surface). The problem of determining the equilibrium magnetic field from external magnetic field measurements has been studied extensively in literature. The problem is 'ill-posed', which means that the solution is unstable to small changes in the measurement data, and the solution has to be constrained in order to stabilize it. Various techniques for handling this problem have been suggested in literature. Usually ad-hoc restrictions are imposed on the equilibrium solution in order to stabilize it. More equilibrium solvers are not able to handle very dissimilar measurement data which means information on the equilibrium is lost. The generally do not allow a straightforward error estimate of the obtained results to be made, and they require large amounts of computing time. This problems are addressed in this thesis. (author). 104 refs.; 42 figs.; 6 tabs
Helical temperature perturbations associated with tearing modes in tokamak plasmas
International Nuclear Information System (INIS)
Fitzpatrick, R.
1994-06-01
An investigation is made into the electron temperature perturbations associated with tearing modes in tokamak plasmas, with a view to determining the mode structure using Electron Cyclotron Emission (ECE) data. It is found that there is a critical magnetic island width below which the conventional picture where the temperature is flattened inside the separatrix is invalid. This effect comes about because of the stagnation of magnetic field lines in the vicinity of the rational surface and the finite parallel thermal conductivity of the plasma. For islands whose widths lie below the critical value there is no flattening of the electron temperature inside the separatrix. Such islands have quite different ECE signatures to conventional magnetic islands. In fact the two island types could, in principle, be differentiated experimentally. It should also be possible to map out the outer ideal magnetohydrodynamical eigenfunctions using ECE data. Islands whose widths are much less than the critical value are not destabilized by the perturbed bootstrap current, unlike conventional magnetic islands. This effect is found to have a number of very interesting consequences and may, indeed, provide an explanation for some puzzling experimental results regarding error field induced magnetic reconnection. All islands whose widths are much greater than the critical width possess a boundary layer on the separatrix which enables heat to be transported from one side of the island to the other via the X-point region. The structure of this boundary layer is described in some detail. Finally, the critical island width is found to be fairly substantial in conventional tokamak plasmas, provided that the long mean free path nature of parallel heat transport and the anomalous nature of perpendicular heat transport are taken into account in the calculation
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
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
Hamiltonian analysis of fast wave current drive in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Becoulet, A; Fraboulet, D; Giruzzi, G; Moreau, D; Saoutic, B [Association Euratom-CEA, Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Chinardet, J [CISI Ingenierie, Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France)
1993-12-01
The Hamiltonian formalism is used to analyze the direct resonant interaction between the fast magnetosonic wave and the electrons in a tokamak plasma. The intrinsic stochasticity of the electron phase space trajectories is derived, and together with extrinsic de-correlation processes, assesses the validity of the quasilinear approximation for the kinetic studies of fast wave current drive (FWCD). A full-wave resolution of the Maxwell-Vlasov set of equations provides the exact pattern of the wave fields in a complete tokamak geometry, for a realistic antenna spectrum. The local quasilinear diffusion tensor is derived from the wave fields, and is used for a computation of the driven current and deposited power profiles, the current drive efficiency, including possible non-linear effects in the kinetic equation. Several applications of FWCD on existing and future machines are given, as well as results concerning combination of FWCD with other non inductive current drive methods. An analytical expression for the current drive efficiency is given in the high single-pass absorption regimes. (authors). 20 figs., 1 tab., 26 refs.
Hamiltonian analysis of fast wave current drive in tokamak plasmas
International Nuclear Information System (INIS)
Becoulet, A.; Fraboulet, D.; Giruzzi, G.; Moreau, D.; Saoutic, B.
1993-12-01
The Hamiltonian formalism is used to analyze the direct resonant interaction between the fast magnetosonic wave and the electrons in a tokamak plasma. The intrinsic stochasticity of the electron phase space trajectories is derived, and together with extrinsic de-correlation processes, assesses the validity of the quasilinear approximation for the kinetic studies of fast wave current drive (FWCD). A full-wave resolution of the Maxwell-Vlasov set of equations provides the exact pattern of the wave fields in a complete tokamak geometry, for a realistic antenna spectrum. The local quasilinear diffusion tensor is derived from the wave fields, and is used for a computation of the driven current and deposited power profiles, the current drive efficiency, including possible non-linear effects in the kinetic equation. Several applications of FWCD on existing and future machines are given, as well as results concerning combination of FWCD with other non inductive current drive methods. An analytical expression for the current drive efficiency is given in the high single-pass absorption regimes. (authors). 20 figs., 1 tab., 26 refs
Modelisation of synchrotron radiation losses in realistic tokamak plasmas
International Nuclear Information System (INIS)
Albajar, F.; Johner, J.; Granata, G.
2000-08-01
Synchrotron radiation losses become significant in the power balance of high-temperature plasmas envisaged for next step tokamaks. Due to the complexity of the exact calculation, these losses are usually roughly estimated with expressions derived from a plasma description using simplifying assumptions on the geometry, radiation absorption, and density and temperature profiles. In the present article, the complete formulation of the transport of synchrotron radiation is performed for realistic conditions of toroidal plasma geometry with elongated cross-section, using an exact method for the calculation of the absorption coefficient, and for arbitrary shapes of density and temperature profiles. The effects of toroidicity and temperature profile on synchrotron radiation losses are analyzed in detail. In particular, when the electron temperature profile is almost flat in the plasma center, as for example in ITB confinement regimes, synchrotron losses are found to be much stronger than in the case where the profile is represented by its best generalized parabolic approximation, though both cases give approximately the same thermal energy contents. Such an effect is not included in present approximate expressions. Finally, we propose a seven-variable fit for the fast calculation of synchrotron radiation losses. This fit is derived from a large database, which has been generated using a code implementing the complete formulation and optimized for massively parallel computing. (author)
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.)
Spectra of heliumlike krypton from tokamak fusion test reactor plasmas
International Nuclear Information System (INIS)
Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M.; Smith, A.; Fraenkel, B.
1993-04-01
Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 Angstrom including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport
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
Theory for neoclassical toroidal plasma viscosity in tokamaks
International Nuclear Information System (INIS)
Shaing, K C; Chu, M S; Hsu, C T; Sabbagh, S A; Seol, Jae Chun; Sun, Y
2012-01-01
Error fields and magnetohydrodynamic modes break toroidal symmetry in tokamaks. The broken symmetry enhances the toroidal plasma viscosity, which results in a steady-state toroidal plasma flow. A theory for neoclassical toroidal plasma viscosity in the low-collisionality regimes is developed. It extends stellarator transport theory to include multiple modes and to allow for |m − nq| ∼ 1. Here, m is the poloidal mode number, n is the toroidal mode number and q is the safety factor. The bounce averaged drift kinetic equation is solved in several asymptotic limits to obtain transport fluxes. These fluxes depend non-linearly on the radial electric field except for those in the 1/ν regime. Here, ν is the collision frequency. The theory is refined to include the effects of the superbanana plateau resonance at the phase space boundary and the finite ∇B drift on the collisional boundary layer fluxes. Analytical expressions that connect all asymptotic limits are constructed and are in good agreement with the numerical results. The flux–force relations that relate transport fluxes to forces are used to illustrate the roles of transport fluxes in the momentum equation. It is shown that the ambipolar state is reached when the momentum equation is relaxed. It is also shown that the origin of the momentum for plasma flow generated without momentum sources is the local unbalance of particles' momenta and is diamagnetic in nature regardless of the details of the theory. (paper)
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
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
Plasma flow and transport on the tokamak ISTTOK boundary plasma
International Nuclear Information System (INIS)
Figueiredo, H.; Silva, C.; Goncalves, B.; Duarte, P.; Fernandes, H.
2011-01-01
The ISTTOK boundary plasma velocity near the outer midplane is measured on the parallel and perpendicular directions in four different configurations by reversing independently the toroidal magnetic field and the plasma current directions. The parallel flow is found to not depend significantly on both the toroidal magnetic field and plasma current directions, being always directed towards the nearest limiter in the scrape-off layer. On the contrary, the perpendicular flow is found to follow the E r x B drift direction. The poloidal velocity has also been derived from the correlation of floating potential signals measured on poloidally separated probes and a good agreement with the value derived with the Gundestrup probe is found. Finally, the dynamical interplay between parallel momentum and turbulent particle flux has been investigated and a clear dynamical coupling between these quantities is found in the region inside the limiter.
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.
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.
Plasma density remote control system of experimental advanced superconductive tokamak
International Nuclear Information System (INIS)
Zhang Mingxin; Luo Jiarong; Li Guiming; Wang Hua; Zhao Dazheng; Xu Congdong
2007-01-01
In Tokamak experiments, experimental data and information on the density control are stored in the local computer system. Therefore, the researchers have to be in the control room for getting the data. Plasma Density Remote Control System (DRCS), which is implemented by encapsulating the business logic on the client in the B/S module, conducts the complicated science computation and realizes the synchronization with the experimental process on the client. At the same time, Web Services and Data File Services are deployed for the data exchange. It is proved in the experiments that DRCS not only meets the requirements for the remote control, but also shows an enhanced capability on the data transmission. (authors)
The Mercier Criterion in Reversed Shear Tokamak Plasmas
International Nuclear Information System (INIS)
Kessel, C.; Chance, M.S.; Jardin, S.C.
1999-01-01
A recent numerical study has found that, contrary to conventional theoretical and experimental expectations, reversed shear plasmas are unstable primarily because the term proportional to the shear in the Mercier criterion is destabilizing. In the present study, the role of the magnetic shear, both local and global, is examined for various tokamak configurations with monotonic and non-monotonic safety factor profiles. The enhancement of the local shear due to the outward shift of the magnetic axis suggests that the latter are less susceptible to interchanges. Furthermore, by regrouping the terms in the criterion, the V'' term when differentiated instead with respect to the toroidal flux, is shown to absorb the dominant shear term. No Mercier instability is found for similar profiles as in the previous study
Simulations of Neon Pellets for Plasma Disruption Mitigation in Tokamaks
Bosviel, Nicolas; Samulyak, Roman; Parks, Paul
2017-10-01
Numerical studies of the ablation of neon pellets in tokamaks in the plasma disruption mitigation parameter space have been performed using a time-dependent pellet ablation model based on the front tracking code FronTier-MHD. The main features of the model include the explicit tracking of the solid pellet/ablated gas interface, a self-consistent evolving potential distribution in the ablation cloud, JxB forces, atomic processes, and an improved electrical conductivity model. The equation of state model accounts for atomic processes in the ablation cloud as well as deviations from the ideal gas law in the dense, cold layers of neon gas near the pellet surface. Simulations predict processes in the ablation cloud and pellet ablation rates and address the sensitivity of pellet ablation processes to details of physics models, in particular the equation of state.
Sustained high βN plasmas on EAST tokamak
Gao, Xiang; the EAST team
2018-05-01
Sustained high normalized beta (βN ∼ 1.9) plasmas with an ITER-like tungsten divertor have been achieved on EAST tokamak recently. The high power NBI heating system of 4.8 MW and the 4.6 GHz lower hybrid wave of 1 MW were developed and applied to produce edge and internal transport barriers in high βN discharges. The central flat q profile with q (ρ) ∼ 1 at ρ safety factor q95 = 4.7 is identified by the multi-channel far-infrared laser polarimeter and the EFIT code. The fraction of non-inductive current is about 40%. The relation between fishbone activity and ITB formation is observed and discussed.
Modeling of noble gas injection into tokamak plasmas
International Nuclear Information System (INIS)
Morozov, D.Kh.; Yurchenko, E.I.; Lukash, V.E.; Baronova, E.O.; Rozhansky, V.A.; Senichenkov, I.Yu.; Veselova, I.Yu.; Schneider, R.
2005-01-01
Noble gas injection for mitigation of the disruption in DIII-D is simulated. The simulation of the first two stages is performed: of the neutral gas jet penetration through the background plasmas, and of the thermal quench. In order to simulate the first stage the 1.5-dimensional numerical code LLP with improved radiation model for noble gas is used. It is demonstrated that the jet remains mainly neutral and thus is able to penetrate to the central region of the tokamak in accordance with experimental observations. Plasma cooling at this stage is provided by the energy exchange with the jet. The radiation is relatively small, and the plasma thermal energy is spent mainly on the jet expansion. The magnetic surfaces in contact with the jet are cooled significantly. The cooling front propagates towards the plasma center. The simulations of the plasma column dynamics in the presence of moving jet is performed by means of the free boundary transport modeling DINA code. It has been shown that the cooling front is accompanied by strongly localized 'shark fin-like' perturbation in toroidal current density profile. After few milliseconds the jet (together with the current perturbation) achieves the region where safety factor is slightly higher than unity and a new type of the non-local kink mode develops. The unstable kink perturbation is non-resonant for any magnetic surface, both inside the plasma column, and in the vacuum space. The mode disturbs mainly the core region. The growth time of the 'shark fin-like' mode is higher than the Alfven time by a factor of 100 for DIII-D parameters. Hence, the simulation describes the DIII-D experimental results, at least, qualitatively. (author)
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.)
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.)
Numerical study of neoclassical plasma pedestal in a tokamak geometry
International Nuclear Information System (INIS)
Chang, C.S.; Ku, Seunghoe; Weitzner, H.
2004-01-01
The fundamental properties of steep neoclassical plasma pedestals in a quiescent tokamak plasma have been investigated with a new guiding center particle code XGC: an X-point included Guiding Center code. It is shown that the width of the steepest neoclassical pedestals is similar to an experimentally observed edge pedestal width, and that a steep pedestal must be accompanied by a self-consistent negative radial electric field well. It is also shown that a steep neoclassical pedestal can form naturally at a quiescent diverted edge as the particle source from the neutral penetration (and heat flux from the core plasma) is balanced by the sharply increasing convective ion loss toward the separatrix. The steep neoclassical pedestal and the strong radial electric field well are suppressed by an anomalous diffusion coefficient of a strength appropriate to an L-mode state; nonetheless, the ExB shearing rate increases rapidly with pedestal temperature. Additionally, the present study shows that a steep pedestal at the diverted edge acts as a cocurrent parallel momentum source
Plasma pressure in the discharge column of the Novillo Tokamak
International Nuclear Information System (INIS)
Gaytan G, E.
1995-01-01
The design and construction of an acquisition system for the measurement of the plasma pressure in the Novillo Tokamak is described in detail. The system includes a high voltage ramp generator, a hardware and a software interface with a personal computer. It is used to determine experimentally the variations of the pressure in the plasma column in the cleaning and main discharges. The measurement of the pressure is made with a Pirani sensor adapted to the acquisition hardware and synchronized with the discharge in the plasma. The software is made in object oriented programming as a graphic interface designed to be used easily. It controls the acquisition, records the data, displays in graphic form the results and save the measurements. The graphic interface is a building block that can be used in different acquisition tasks. The ramp generator can deliver a signal of 200 V peak to peak with a current of 200 m A and offset control. The acquisition time is 2.5 μ s for every measurement, 8192 measurements can be stored in the acquisition board for every discharge. (Author)
Analysis of IBW-driven plasma flows in tokamaks
International Nuclear Information System (INIS)
Berry, L.A.; Jaeger, E.F.; D'Azevedo, E.F.; Batchelor, D.B.; Carlsson, J.A.; Carter, M.D.; Cesario, R.
2001-01-01
Both theory and experiment have suggested that damping of Ion Bernstein Waves (IBWs) at ion cyclotron frequency harmonics could drive poloidal flows and lead to enhanced confinement for tokamaks. However, the early analyses were based on Reynolds stress closures of moment equations. More rigorous, finite Larmor radius (FLR) expansions of the radio frequency (RF) kinetic pressure for low harmonic interactions indicated that the Reynolds stress approximation was not generally valid, and resulted in significant changes in the plasma flow response. These changes were largest for wave interactions driven by finite Larmour radius effects. To provide a better assessment of higher harmonic interactions and IBW flow drive prospects, the electromagnetic (E and M) and RF kinetic force models are extended with no assumptions regarding the smallness of the ion Larmor radius. For both models, a spectral-width approximation was used to make the numerical analysis tractable. In addition, it was necessary to include the effects of plasma equilibrium gradients on the plasma conductivity and the RF-induced momentum in order to conserve energy and momentum. The analysis of high-harmonic IBW interactions for TFTR and FTU parameters indicates significant poloidal flow shears (relative to turbulence correlation times) for power levels available in present experiments. Recent advances in all-orders calculations of E and M fields in 2-D are also discussed. (author)
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)
Perturbative transport experiments in JET Advanced Tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Mantica, P.; Gorini, G.; Sozzi, C. [Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Association, Milan (Italy); Imbeaux, F.; Sarazin, Y.; Garbet, X. [Association Euratom-CEA, St. Paul-lez-Durance Cedex (France); Kinsey, J. [Lehigh Univ., Bethlehem, Pennsylvania (United States); Budny, R. [Princeton Plasma Physics Lab, New Jersey (United States); Coffey, I.; Parail, V.; Walden, A. [Euratom/UKAEA Fusion Association, Abingdon, Oxon (United Kingdom); Dux, R. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Garzotti, L. [Istituto Gas Ionizzati, Padova (Italy); Ingesson, C. [FOM-Instituut voor Plasmafysica, Nieuwegein (Netherlands); Kissick, M. [University of California, Los Angeles (United States)
2003-07-01
Perturbative transport experiments have been performed in JET Advanced Tokamak plasmas either in conditions of fully developed Internal Transport Barrier (ITB) or during a phase where an ITB was not observed. Transient peripheral cooling was induced by either Laser Ablation or Shallow Pellet Injection and the ensuing travelling cold pulse was used to probe the plasma transport in the electron and, for the first time, also in the ion channel. Cold pulses travelling through ITBs are observed to erode the ITB outer part, but, if the inner ITB portion survives, it strongly damps the propagating wave. The result is discussed in the context of proposed possible pictures for ITB formation. In the absence of an ITB, the cold pulse shows a fast propagation in the outer plasma half, which is consistent with a region of stiff transport, while in the inner half it slows down but shows the peculiar feature of amplitude growing while propagating. The data are powerful tests for the validation of theoretical transport models. (author)
Internal transport barrier in tokamak and helical plasmas
Ida, K.; Fujita, T.
2018-03-01
The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, R/{L}T, the location, {ρ }{ITB}, and the width, W/a, and can be expressed by ‘weak’ ITB (small R/{L}T) or ‘strong’ (large R/{L}T), ‘small’ ITB (small {ρ }{ITB}) or ‘large’ ITB (large {ρ }{ITB}), and ‘narrow’ (small W/a) or ‘wide’ (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the
Czech Academy of Sciences Publication Activity Database
Svoboda, V.; Kocman, J.; Grover, O.; Krbec, Jaroslav; Stöckel, Jan
96-97, October (2015), s. 974-979 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] Institutional support: RVO:61389021 Keywords : tokamak technology * remote participation * plasma stabilization Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://dx.doi.org/10.1016/j.fusengdes.2015.06.044
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
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
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
High density plasma heating in the Tokamak à configuration variable
International Nuclear Information System (INIS)
Curchod, L.
2011-04-01
The Tokamak à Configuration Variable (TCV) is a medium size magnetic confinement thermonuclear fusion experiment designed for the study of the plasma performances as a function of its shape. It is equipped with a high power and highly flexible electron cyclotron heating (ECH) and current drive (ECCD) system. Up to 3 MW of 2 nd harmonic EC power in ordinary (O 2 ) or extraordinary (X 2 ) polarization can be injected from TCV low-field side via six independently steerable launchers. In addition, up to 1.5 MW of 3 rd harmonic EC power (X 3 ) can be launched along the EC resonance from the top of TCV vacuum vessel. At high density, standard ECH and ECCD are prevented by the appearance of a cutoff layer screening the access to the EC resonance at the plasma center. As a consequence, less than 50% of TCV density operational domain is accessible to X 2 and X 3 ECH. The electron Bernstein waves (EBW) have been proposed to overcome this limitation. EBW is an electrostatic mode propagating beyond the plasma cutoff without upper density limit. Since it cannot propagate in vacuum, it has to be excited by mode conversion of EC waves in the plasma. Efficient electron Bernstein waves heating (EBH) and current drive (EBCD) were previously performed in several fusion devices, in particular in the W7-AS stellarator and in the MAST spherical tokamak. In TCV, the conditions for an efficient O-X-B mode conversion (i.e. a steep density gradient at the O 2 plasma cutoff) are met at the edge of high confinement (H-mode) plasmas characterized by the appearance of a pedestal in the electron temperature and density profiles. TCV experiments have demonstrated the first EBW coupling to overdense plasmas in a medium aspect-ratio tokamak via O-X-B mode conversion. This thesis work focuses on several aspects of ECH and EBH in low and high density plasmas. Firstly, the experimental optimum angles for the O-X-B mode conversion is successfully compared to the full-wave mode conversion calculation
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
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
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
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
Anomalous electron streaming due to electrostatic modes in tokamak plasmas
International Nuclear Information System (INIS)
Schultz, S.D.; Bers, A.; Ram, A.K.
1993-01-01
The motion of circulating electrons in a tokamak interacting with electrostatic waves (such as lower-hybrid waves) is given by a guiding center Hamiltonian and studied by numerical integration. The unperturbed motion of electron guiding centers is first shown to be integrable, and, in a manner similar to that used in previous works, a set of action-angle coordinates for the orbits are derived which take into account finite aspect ratio and noncircular plasma cross section. Electrostatic modes in the low-frequency, long-wavelength limit are treated as a perturbation to the guiding center Hamiltonian. The waves are generated with low integral values of the toroidal and poloidal mode numbers n and m and satisfy the approximate lower-hybrid dispersion relation k perpendicular /k parallel ∼ ω pe /ω ∼ 10 1.5 . If the number of modes is greater than three, the electron motion parallel to the magnetic field is observed to be stochastic in the phase-space region where v parallel is near the wave parallel phase velocity. On surfaces with rational values of the safety factor q, superposition of modes with degenerate values of the parallel mode number n + (m/q) is shown to result in electron streaming perpendicular to the magnetic field. The speed and direction of this radial motion are observed to have sinusoidal dependence on the poloidal angle. For models including finite magnetic-field shear, the authors find a limit to the extent of the radial streaming of the electrons. Results for the speed of the electron radial motion for typical tokamak parameters are presented
Kinetic modelling of runaway electron avalanches in tokamak plasmas
International Nuclear Information System (INIS)
Nilsson, E; Peysson, Y; Saint-Laurent, F; Decker, J; Granetz, R S; Vlainic, M
2015-01-01
Runaway electrons can be generated in tokamak plasmas if the accelerating force from the toroidal electric field exceeds the collisional drag force owing to Coulomb collisions with the background plasma. In ITER, disruptions are expected to generate runaway electrons mainly through knock-on collisions (Hender et al 2007 Nucl. Fusion 47 S128–202), where enough momentum can be transferred from existing runaways to slow electrons to transport the latter beyond a critical momentum, setting off an avalanche of runaway electrons. Since knock-on runaways are usually scattered off with a significant perpendicular component of the momentum with respect to the local magnetic field direction, these particles are highly magnetized. Consequently, the momentum dynamics require a full 3D kinetic description, since these electrons are highly sensitive to the magnetic non-uniformity of a toroidal configuration. For this purpose, a bounce-averaged knock-on source term is derived. The generation of runaway electrons from the combined effect of Dreicer mechanism and knock-on collision process is studied with the code LUKE, a solver of the 3D linearized bounce-averaged relativistic electron Fokker–Planck equation (Decker and Peysson 2004 DKE: a fast numerical solver for the 3D drift kinetic equation Report EUR-CEA-FC-1736, Euratom-CEA), through the calculation of the response of the electron distribution function to a constant parallel electric field. The model, which has been successfully benchmarked against the standard Dreicer runaway theory now describes the runaway generation by knock-on collisions as proposed by Rosenbluth (Rosenbluth and Putvinski 1997 Nucl. Fusion 37 1355–62). This paper shows that the avalanche effect can be important even in non-disruptive scenarios. Runaway formation through knock-on collisions is found to be strongly reduced when taking place off the magnetic axis, since trapped electrons can not contribute to the runaway electron population. Finally
Nonlinear transport processes in tokamak plasmas. I. The collisional regimes
International Nuclear Information System (INIS)
Sonnino, Giorgio; Peeters, Philippe
2008-01-01
An application of the thermodynamic field theory (TFT) to transport processes in L-mode tokamak plasmas is presented. The nonlinear corrections to the linear ('Onsager') transport coefficients in the collisional regimes are derived. A quite encouraging result is the appearance of an asymmetry between the Pfirsch-Schlueter (P-S) ion and electron transport coefficients: the latter presents a nonlinear correction, which is absent for the ions, and makes the radial electron coefficients much larger than the former. Explicit calculations and comparisons between the neoclassical results and the TFT predictions for Joint European Torus (JET) plasmas are also reported. It is found that the nonlinear electron P-S transport coefficients exceed the values provided by neoclassical theory by a factor that may be of the order 10 2 . The nonlinear classical coefficients exceed the neoclassical ones by a factor that may be of order 2. For JET, the discrepancy between experimental and theoretical results for the electron losses is therefore significantly reduced by a factor 10 2 when the nonlinear contributions are duly taken into account but, there is still a factor of 10 2 to be explained. This is most likely due to turbulence. The expressions of the ion transport coefficients, determined by the neoclassical theory in these two regimes, remain unaltered. The low-collisional regimes, i.e., the plateau and the banana regimes, are analyzed in the second part of this work
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.)
Double internal transport barrier triggering mechanism in tokamak plasmas
International Nuclear Information System (INIS)
Dong, Jiaqi; Mou, Zongze; Long, Yongxing; Mahajan, Swadesh M.
2004-01-01
Sheared flow layers created by energy released in magnetic reconnection processes are studied with the magneto hydrodynamics (MHD), aimed at internal transport barrier (ITB) dynamics. The double tearing mode induced by electron viscosity is investigated and proposed as a triggering mechanism for double internal transport barrier (DITB) observed in tokamak plasmas with non-monotonic safety factor profiles. The quasi-linear development of the mode is simulated and the emphasis is placed on the structure of sheared poloidal flow layers formed in the vicinity of the magnetic islands. For viscosity double tearing modes, it is shown that the sheared flows induced by the mode may reach the level required by the condition for ITB formation. Especially, the flow layers are found to form just outside the magnetic islands. The scaling of the generated velocity with plasma parameters is given. Possible explanation for the experimental observations that the preferential formation of transport barriers in the proximity of low order rational surface is discussed. (author)
Plasma rotation and transport in MAST spherical tokamak
Field, A. R.; Michael, C.; Akers, R. J.; Candy, J.; Colyer, G.; Guttenfelder, W.; Ghim, Y.-c.; Roach, C. M.; Saarelma, S.; MAST Team
2011-06-01
The formation of internal transport barriers (ITBs) is investigated in MAST spherical tokamak plasmas. The relative importance of equilibrium flow shear and magnetic shear in their formation and evolution is investigated using data from high-resolution kinetic- and q-profile diagnostics. In L-mode plasmas, with co-current directed NBI heating, ITBs in the momentum and ion thermal channels form in the negative shear region just inside qmin. In the ITB region the anomalous ion thermal transport is suppressed, with ion thermal transport close to the neo-classical level, although the electron transport remains anomalous. Linear stability analysis with the gyro-kinetic code GS2 shows that all electrostatic micro-instabilities are stable in the negative magnetic shear region in the core, both with and without flow shear. Outside the ITB, in the region of positive magnetic shear and relatively weak flow shear, electrostatic micro-instabilities become unstable over a wide range of wave numbers. Flow shear reduces the linear growth rates of low-k modes but suppression of ITG modes is incomplete, which is consistent with the observed anomalous ion transport in this region; however, flow shear has little impact on growth rates of high-k, electron-scale modes. With counter-NBI ITBs of greater radial extent form outside qmin due to the broader profile of E × B flow shear produced by the greater prompt fast-ion loss torque.
Plasma control issues for an advanced steady state tokamak reactor
International Nuclear Information System (INIS)
Moreau, D.
2001-01-01
This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)
Plasma control issues for an advanced steady state tokamak reactor
International Nuclear Information System (INIS)
Moreau, D.; Voitsekhovitch, I.
1999-01-01
This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)
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)
Dynamics of the edge transport barrier at plasma biasing on the CASTOR tokamak
Czech Academy of Sciences Publication Activity Database
Stöckel, Jan; Spolaore, M.; Peleman, P.; Brotánková, Jana; Horáček, Jan; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Kocan, M.; Martines, E.; Pánek, Radomír; Sharma, A.; Van Oost, G.
2006-01-01
Roč. 12, č. 6 (2006), s. 19-23 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * transport barrier * relaxations Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html
Czech Academy of Sciences Publication Activity Database
Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Krämer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Peleman, P.; Razumova, K.; Stöckel, Jan; Vershkov, V.; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Dnestrovskij, A.Yu.; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Jachmin, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Soldatov, S.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Zimmermann, O.
2006-01-01
Roč. 12, č. 6 (2006), s. 14-19 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * improved confinement * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html
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)
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
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.
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.)
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.).
High-Q plasmas in the TFTR tokamak
International Nuclear Information System (INIS)
Jassby, D.L.; Barnes, C.W.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Dylla, H.F.; Efthimion, P.C.; Fredrickson, E.D.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.C.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S.J.; LaMarche, P.H.; LeBlanc, B.; Mansfield, D.K.; Marmar, E.S.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.K.; Paul, S.F.; Pitcher, S.; Ramsey, A.T.; Redi, M.H.; Sabbagh, S.A.; Scott, S.D.; Snipes, J.; Stevens, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Terry, J.L.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.
1991-01-01
In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength S n and D--D fusion power gain Q DD are realized in the neutral-beam-fueled and heated ''supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, S n increases approximately as P 1.8 b . The highest-Q shots are characterized by high T e (up to 12 keV), T i (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad T e profiles, and lower Z eff . Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the ''carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, Q DD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness [n e (0)/left-angle n e right-angle] during the beam pulse. To date, the best fusion results are S n =5x10 16 n/sec, Q DD =1.85x10 -3 , and neutron yield=4.0x10 16 n/pulse, obtained at I p =1.6--1.9 MA and beam energy E b =95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of S n arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with P b
Optimization of tokamak plasma equilibrium shape using parallel genetic algorithms
International Nuclear Information System (INIS)
Zhulin An; Bin Wu; Lijian Qiu
2006-01-01
In the device of non-circular cross sectional tokamaks, the plasma equilibrium shape has a strong influence on the confinement and MHD stability. The plasma equilibrium shape is determined by the configuration of the poloidal field (PF) system. Usually there are many PF systems that could support the specified plasma equilibrium, the differences are the number of coils used, their positions, sizes and currents. It is necessary to find the optimal choice that meets the engineering constrains, which is often done by a constrained optimization. The Genetic Algorithms (GAs) based method has been used to solve the problem of the optimization, but the time complexity limits the algorithms to become widely used. Due to the large search space that the optimization has, it takes several hours to get a nice result. The inherent parallelism in GAs can be exploited to enhance their search efficiency. In this paper, we introduce a parallel genetic algorithms (PGAs) based approach which can reduce the computational time. The algorithm has a master-slave structure, the slave explore the search space separately and return the results to the master. A program is also developed, and it can be running on any computers which support massage passing interface. Both the algorithm and the program are detailed discussed in the paper. We also include an application that uses the program to determine the positions and currents of PF coils in EAST. The program reach the target value within half an hour and yield a speedup rate of 5.21 on 8 CPUs. (author)
Electromagnetic microinstabilities in tokamak plasmas using a global spectral approach
Energy Technology Data Exchange (ETDEWEB)
Falchetto, G. L
2002-03-01
Electromagnetic microinstabilities in tokamak plasmas are studied by means of a linear global eigenvalue numerical code. The code is the electromagnetic extension of an existing electrostatic global gyrokinetic spectral toroidal code, called GLOGYSTO. Ion dynamics is described by the gyrokinetic equation, so that ion finite Larmor radius effects are taken into account to all orders. Non adiabatic electrons are included in the model, with passing particles described by the drift-kinetic equation and trapped particles through the bounce averaged drift-kinetic equation. A low frequency electromagnetic perturbation is applied to a low -but finite- {beta}plasma (where the parameter {beta} identifies the ratio of plasma pressure to magnetic pressure); thus, the parallel perturbations of the magnetic field are neglected. The system is closed by the quasi-neutrality equation and the parallel component of Ampere's law. The formulation is applied to a large aspect ratio toroidal configuration, with circular shifted surfaces. Such a simple configuration enables one to derive analytically the gyrocenter trajectories. The system is solved in Fourier space, taking advantage of a decomposition adapted to the toroidal geometry. The major contributions of this thesis are as follows. The electromagnetic effects on toroidal Ion Temperature Gradient driven (ITG) modes are studied. The stabilization of these modes with increasing {beta}, as predicted in previous work, is confirmed. The inclusion of trapped electron dynamics enables the study of its coupling to the ITG modes and of Trapped Electron Modes (TEM) .The effects of finite {beta} are considered together with those of different magnetic shear profiles and of the Shafranov shift. The threshold for the destabilization of an electromagnetic mode is identified. Moreover, the global formulation yields for the first time the radial structure of this so-called Alfvenic Ion Temperature Gradient (AITG) mode. The stability of the
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
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.
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
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)
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.)
Two-fluid and parallel compressibility effects in tokamak plasmas
International Nuclear Information System (INIS)
Sugiyama, L.E.; Park, W.
1998-01-01
The MHD, or single fluid, model for a plasma has long been known to provide a surprisingly good description of much of the observed nonlinear dynamics of confined plasmas, considering its simple nature compared to the complexity of the real system. On the other hand, some of the supposed agreement arises from the lack of the detailed measurements that are needed to distinguish MHD from more sophisticated models that incorporate slower time scale processes. At present, a number of factors combine to make models beyond MHD of practical interest. Computational considerations still favor fluid rather than particle models for description of the full plasma, and suggest an approach that starts from a set of fluid-like equations that extends MHD to slower time scales and more accurate parallel dynamics. This paper summarizes a set of two-fluid equations for toroidal (tokamak) geometry that has been developed and tested as the MH3D-T code [1] and some results from the model. The electrons and ions are described as separate fluids. The code and its original MHD version, MH3D [2], are the first numerical, initial value models in toroidal geometry that include the full 3D (fluid) compressibility and electromagnetic effects. Previous nonlinear MHD codes for toroidal geometry have, in practice, neglected the plasma density evolution, on the grounds that MHD plasmas are only weakly compressible and that the background density variation is weaker than the temperature variation. Analytically, the common use of toroidal plasma models based on aspect ratio expansion, such as reduced MHD, has reinforced this impression, since this ordering reduces plasma compressibility effects. For two-fluid plasmas, the density evolution cannot be neglected in principle, since it provides the basic driving energy for the diamagnetic drifts of the electrons and ions perpendicular to the magnetic field. It also strongly influences the parallel dynamics, in combination with the parallel thermal
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)
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
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
Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST
International Nuclear Information System (INIS)
Xu, X Q
2007-01-01
We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D (ψ, θ, ε, μ) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices
Locked magnetic island chains in toroidally flow damped tokamak plasmas
International Nuclear Information System (INIS)
Fitzpatrick, R; Waelbroeck, F L
2010-01-01
The physics of a locked magnetic island chain maintained in the pedestal of an H-mode tokamak plasma by a static, externally generated, multi-harmonic, helical magnetic perturbation is investigated. The non-resonant harmonics of the external perturbation are assumed to give rise to significant toroidal flow damping in the pedestal, in addition to the naturally occurring poloidal flow damping. Furthermore, the flow damping is assumed to be sufficiently strong to relax the pedestal ion toroidal and poloidal fluid velocities to fixed values determined by neoclassical theory. The resulting neoclassical ion flow causes a helical phase-shift to develop between the locked island chain and the resonant harmonic of the external perturbation. Furthermore, when this phase-shift exceeds a critical value, the chain unlocks from the resonant harmonic and starts to rotate, after which it decays away and is replaced by a helical current sheet. The neoclassical flow also generates an ion polarization current in the vicinity of the island chain which either increases or decreases the chain's radial width, depending on the direction of the flow. If the polarization effect is stabilizing, and exceeds a critical amplitude, then the helical island equilibrium becomes unstable, and the chain again decays away. The critical amplitude of the resonant harmonic of the external perturbation at which the island chain either unlocks or becomes unstable is calculated as a function of the pedestal ion pressure, the neoclassical poloidal and toroidal ion velocities and the poloidal and toroidal flow damping rates.
Plasma driving system requirements for commercial tokamak fusion reactors
International Nuclear Information System (INIS)
Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.
1978-01-01
The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration
Plasma driving system requirements for commercial tokamak fusion reactors
International Nuclear Information System (INIS)
Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.
1977-01-01
The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration
Understanding L-H transition in tokamak fusion plasmas
Xu, Guosheng; Wu, Xingquan
2017-03-01
This paper reviews the current state of understanding of the L-H transition phenomenon in tokamak plasmas with a focus on two central issues: (a) the mechanism for turbulence quick suppression at the L-H transition; (b) the mechanism for subsequent generation of sheared flow. We briefly review recent advances in the understanding of the fast suppression of edge turbulence across the L-H transition. We uncover a comprehensive physical picture of the L-H transition by piecing together a number of recent experimental observations and insights obtained from 1D and 2D simulation models. Different roles played by diamagnetic mean flow, neoclassical-driven mean flow, turbulence-driven mean flow, and turbulence-driven zonal flows are discussed and clarified. It is found that the L-H transition occurs spontaneously mediated by a shift in the radial wavenumber spectrum of edge turbulence, which provides a critical evidence for the theory of turbulence quench by the flow shear. Remaining questions and some key directions for future investigations are proposed. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contracts No. 2015GB101000, No. 2013GB106000, and No. 2013GB107000 and National Natural Science Foundation of China under Contracts No. 11575235 and No. 11422546.
Electron Landau damping of ion Bernstein waves in tokamak plasmas
International Nuclear Information System (INIS)
Brambilla, M.
1998-01-01
Absorption of ion Bernstein (IB) waves by electrons is investigated. These waves are excited by linear mode conversion in tokamak plasmas during fast wave (FW) heating and current drive experiments in the ion cyclotron range of frequencies. Near mode conversion, electromagnetic corrections to the local dispersion relation largely suppress electron Landau damping of these waves, which becomes important again, however, when their wavelength is comparable to the ion Larmor radius or shorter. The small Larmor radius wave equations solved by most numerical codes do not correctly describe the onset of electron Landau damping at very short wavelengths, and these codes, therefore, predict very little damping of IB waves, in contrast to what one would expect from the local dispersion relation. We present a heuristic, but quantitatively accurate, model which allows account to be taken of electron Landau damping of IB waves in such codes, without affecting the damping of the compressional wave or the efficiency of mode conversion. The possibilities and limitations of this approach are discussed on the basis of a few examples, obtained by implementing this model in the toroidal axisymmetric full wave code TORIC. (author)
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
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.)
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
International Nuclear Information System (INIS)
Chen, S Y; Hong, B B; Liu, Y; Lu, W; Huang, J; Tang, C J; Ding, X T; Zhang, X J; Hu, Y J
2012-01-01
The synergy between electron cyclotron current drive (ECCD) and lower hybrid current drive (LHCD) is investigated numerically with the parameters of the HL-2A tokamak. Based on the understanding of the synergy mechanisms, a high current driven efficiency or a desired radial current profile can be achieved through properly matching the parameters of ECCD and LHCD due to the flexibility of ECCD. Meanwhile, it is found that the total current driven by the electron cyclotron wave (ECW) and the lower hybrid wave (LHW) simultaneously can be smaller than the sum of the currents driven by the ECW and LHW separately, when the power of the ECW is much larger than the LHW power. One of the reasons leading to this phenomenon (referred to as negative synergy in this context) is that fast current-carrying electrons tend to be trapped, when the perpendicular velocity driven by the ECW is large and the parallel velocity decided by the LHW is correspondingly small. (paper)
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
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)
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.
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
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
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
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
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
Energy Technology Data Exchange (ETDEWEB)
Nakamura, Y; Watanabe, T; Nagao, A; Nakamura, K; Kikuchi, M; Aoki, T; Hiraki, N; Itoh, S [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics; Mitarai, O
1982-02-01
Critical condition for current-driven instability excited in turbulently heated TRIAM-1 tokamak plasma is investigated experimentally. Resistive hump in loop voltage, plasma density fluctuation and rapid increase of electron temperature in a skin layer are simultaneously observed at the time when the electron drift velocity amounts to the critical drift velocity for low-frequency ion acoustic instability.
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
Witvoet, G.; Westerhof, E.; Steinbuch, M.; Doelman, N.J.; Baar, de M.R.
2009-01-01
Tokamak plasmas in nuclear fusion are subject to various instabilities. A clear example is the sawtooth instability, which has both positive and negative effects on the plasma. To optimize between these effects control of the sawtooth period is necessary. This paper presents a simple control
Electron cyclotron absorption in Tokamak plasmas in the presence of radial transport of particles
International Nuclear Information System (INIS)
Rosa, Paulo R. da S.; Ziebell, Luiz F.
1998-01-01
We use quasilinear theory to study effects of particle radial transport on the electron cyclotron absorption coefficient by a current carrying plasma, in a tokamak modelated as a plasma slab. Our numerical results indicate significant modification in the profile of the electron cyclotron absorption coefficient when transport is taken into account relative to the situation without transport. (author)
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)
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)
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)
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
Study of Globus-M Tokamak Poloidal System and Plasma Position Control
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.
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)
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)
International Nuclear Information System (INIS)
Murphy, T.J.
1986-11-01
In a recently proposed positron transport experiment, positrons would be deposited in a fusion plasma by forming a positronium (Ps) beam and passing it through the plasma. Positrons would be deposited as the beam is ionized by plasma ions and electrons. Radial transport of the positrons to the limiter could then be measured by detecting the gamma radiation produced by annihilation of positrons with electrons in the limiter. This would allow measurements of the transport of electron-mass particles and might shed some light on the mechanisms of electron transport in fusion plasmas. In this paper, the deposition and transport of positrons in a tokamak are simulated and the annihilation signal determined for several transport models. Calculations of the expected signals are necessary for the optimal design of a positron transport experiment. There are several mechanisms for the loss of positrons besides transport to the limiter. Annihilation with plasma electrons and reformation of positronium in positron-hydrogen collisions are two such processes. These processes can alter the signal and place restrictions ons on the plasma conditions in which positron transport experiments can be effectively performed
Impedance of an intense plasma-cathode electron source for tokamak startup
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.
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)
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
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
Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor
International Nuclear Information System (INIS)
Efthimion, P.C.; Goeler, S. von; Houlberg, W.A.
2001-01-01
Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)
Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor
International Nuclear Information System (INIS)
Efthimion, P.C.; Von Goeler, S.; Houlberg, W.A.
1999-01-01
Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)
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)
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.
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
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
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
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)
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
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)
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.
Magnetic evaluation of hydrogen pressures changes on MHD fluctuations in IR-T1 tokamak plasma
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.
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
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)
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
Large potential change induced by pellet injection in JIPP T-IIU tokamak plasmas
International Nuclear Information System (INIS)
Hamada, Y.; Sato, K.N.; Sakakita, H.
1995-05-01
A large, rapid change in the local plasma potential is found to be induced by off-axis hydrogen ice-pellet injection into a tokamak plasma. The polarity of the rapid change is reversed when the pellet is injected into the upper and lower halves of the poloidal plasma cross-section. This change can be interpreted as being due to the gradient-B drift of particles in the high-density plasmas of the pellet cloud, before the increase of the plasma density due to the ablation becomes uniform on the magnetic surface. (author)
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
Diagnostic applications of transient synchrotron radiation in tokamak plasmas
International Nuclear Information System (INIS)
Fisch, N.J.; Kritz, A.H.
1990-02-01
Transient radiation, resulting from a brief, deliberate perturbation of the velocity distribution of superthermal tokamak electrons, can be more informative than the steady background radiation that is present in the absence of the perturbation. It is possible to define a number of interesting inverse problems, which exploit the two-dimensional frequency-time data of the transient radiation signal. 17 refs
Analysis of tokamak plasma confinement modes using the fast ...
Indian Academy of Sciences (India)
2016-10-20
Oct 20, 2016 ... ruby laser technology have been extended to deter- mine q. The edge radial electric field is known to ... The control of the shear layer is therefore important to modify transport in tokamaks [26–33]. A ... The calculation of PSD is directly accomplished by the FFT technique presented in the MATLAB software.
Quasilinear kinetic modeling of RMP penetration into a tokamak plasma
International Nuclear Information System (INIS)
Heyn, M.F.; Kernbichler, W.; Leitner, P.; Ivanov, I.B.; Kasilov, S.V.
2013-01-01
The linear as well as the quasilinear problem of RMP penetration in tokamaks is solved consistently with a particle and energy conserving collision operator. The new collision operator ensures the Onsager symmetry of the quasilinear transport coefficient matrix and avoids artifacts such as fake heat convection connected with simplified collision models.
Strong Scattering of High Power Millimeter Waves in Tokamak Plasmas with Tearing Modes
DEFF Research Database (Denmark)
Westerhof, E.; Nielsen, Stefan Kragh; Oosterbeek, J.W.
2009-01-01
In tokamak plasmas with a tearing mode, strong scattering of high power millimeter waves, as used for heating and noninductive current drive, is shown to occur. This new wave scattering phenomenon is shown to be related to the passage of the O point of a magnetic island through the high power...
Effects of radial envelope modulations on the collisionless trapped-electron mode in tokamak plasmas
Chen, Hao-Tian; Chen, Liu
2018-05-01
Adopting the ballooning-mode representation and including the effects of radial envelope modulations, we have derived the corresponding linear eigenmode equation for the collisionless trapped-electron mode in tokamak plasmas. Numerical solutions of the eigenmode equation indicate that finite radial envelope modulations can affect the linear stability properties both quantitatively and qualitatively via the significant modifications in the corresponding eigenmode structures.
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
Meyer, H.; Eich, T.; Beurskens, M.N.A.; Coda, S.; Hakola, A.; Martin, P.; Adamek, J.; Agostini, M.; Aguiam, D.; Ahn, J.; Aho-Mantila, L.; Akers, R.; Albanese, R.; Aledda, R.; Alessi, E.; Allan, S.; Alves, D.; Ambrosino, R.; Amicucci, L.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Apruzzese, G.; Ariola, M.; Arnichand, H.; Arter, W.; Baciero, A.; Barnes, M.; Barrera, L.; Behn, R.; Bencze, A.; Bernardo, J.; Bernert, M.; Bettini, P.; Bilková, P.; Bin, W.; Birkenmeier, G.; Bizarro, J. P.S.; Blanchard, P.; Blanken, T.; Bluteau, M.; Bobkov, V.; Bogar, O.; Böhm, P.; Bolzonella, T.; Boncagni, L.; Botrugno, A.; Bottereau, C.; Bouquey, F.; Bourdelle, C.; Brémond, S.; Brezinsek, S.; Brida, D.; Brochard, F.; Buchanan, J.; Bufferand, H.; Buratti, P.; Cahyna, P.; Calabrò, G.; Camenen, Y.; Caniello, R.; Cannas, B.; Canton, A.; Cardinali, A.; Carnevale, D.; Carr, M.; Carralero, D.; Carvalho, P.; Casali, L.; Castaldo, C.; Castejón, F.; Castro, R.; Causa, F.; Cavazzana, R.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cesario, R.; Challis, C.D.; Chapman, I.T.; Chapman, S.; Chernyshova, M.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Clairet, F.; Classen, I.; Coelho, R.; Coenen, J. W.; Colas, L.; Conway, G.; Corre, Y.; Costea, S.; Crisanti, F.; Cruz, N.; Cseh, G.; Czarnecka, A.; D'Arcangelo, O.; De Angeli, M.; De Masi, G.; De Temmerman, G.; De Tommasi, G.; Decker, J.; Delogu, R. S.; Dendy, R.; Denner, P.; Di Troia, C.; Dimitrova, M.; D'Inca, R.; Dorić, V.; Douai, D.; Drenik, A.; Dudson, B.; Dunai, D.; Dunne, M.; Duval, B. P.; Easy, L.; Elmore, S.; Erdös, B.; Esposito, B.; Fable, E.; Faitsch, M.; Fanni, A.; Fedorczak, N.; Felici, F.; Ferreira, J.; Février, O.; Ficker, O.; Fietz, S.; Figini, L.; Figueiredo, A.; Fil, A.; Fishpool, G.; Fitzgerald, M.; Fontana, M.; Ford, O.; Frassinetti, L.; Fridström, R.; Frigione, D.; Fuchert, G.; Fuchs, C.; Furno Palumbo, M.; Futatani, S.; Gabellieri, L.; Gałazka, K.; Galdon-Quiroga, J.; Galeani, S.; Gallart, D.; Gallo, A.; Galperti, C.; Gao, Y.; Garavaglia, S.; Garcia, J.; Garcia-Carrasco, A.; Garcia-Lopez, J.; Garcia-Munoz, M.; Gardarein, J. L.; Garzotti, L.; Gaspar, J.; Gauthier, E.; Geelen, P.; Geiger, B.; Ghendrih, P.; Ghezzi, F.; Giacomelli, L.; Giannone, L.; Giovannozzi, E.; Giroud, C.; Gleason González, C.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Gruber, M.; Gude, A.; Guimarais, L.; Guirlet, R.; Gunn, J.; Hacek, P.; Hacquin, S.; Hall, S.; Ham, C.; Happel, T.; Harrison, J.; Harting, D.; Hauer, V.; Havlickova, E.; Hellsten, T.; Helou, W.; Henderson, S.; Hennequin, P.; Heyn, M.; Hnat, B.; Hölzl, M.; Hogeweij, D.; Honoré, C.; Hopf, C.; Horáček, J.; Hornung, G.; Horváth, L.; Huang, Z.; Huber, A.; Igitkhanov, J.; Igochine, V.; Imrisek, M.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A. S.; Jacquet, P.; Jakubowski, M.; Jardin, A.; Jaulmes, F.; Jenko, F.; Jensen, T.; Jeppe Miki Busk, O.; Jessen, M.; Joffrin, E.; Jones, O.; Jonsson, T.; Kallenbach, A.; Kallinikos, N.; Kálvin, S.; Kappatou, A.; Karhunen, J.; Karpushov, A.; Kasilov, S.; Kasprowicz, G.; Kendl, A.; Kernbichler, W.; Kim, D.; Kirk, A.; Kjer, S.; Klimek, I.; Kocsis, G.; Kogut, D.; Komm, M.; Korsholm, S. B.; Koslowski, H. R.; Koubiti, M.; Kovacic, J.; Kovarik, K.; Krawczyk, N.; Krbec, J.; Krieger, K.; Krivska, A.; Kube, R.; Kudlacek, O.; Kurki-Suonio, T.; Labit, B.; Laggner, F. M.; Laguardia, L.; Lahtinen, A.; Lalousis, P.; Lang, P.; Lauber, P.; Lazányi, N.; Lazaros, A.; Le, H.B.; Lebschy, A.; Leddy, J.; Lefévre, L.; Lehnen, M.; Leipold, F.; Lessig, A.; Leyland, M.; Li, L.; Liang, Y.; Lipschultz, B.; Liu, Y.Q.; Loarer, T.; Loarte, A.; Loewenhoff, T.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Lupelli, I.; Lux, H.; Lyssoivan, A.; Madsen, J.; Maget, P.; Maggi, C.; Maggiora, R.; Magnussen, M. L.; Mailloux, J.; Maljaars, B.; Malygin, A.; Mantica, P.; Mantsinen, M.; Maraschek, M.; Marchand, B.; Marconato, N.; Marini, C.; Marinucci, M.; Markovic, T.; Marocco, D.; Marrelli, L.; Martin, Y.; Martin Solis, J. R.; Martitsch, A.; Mastrostefano, S.; Mattei, M.; Matthews, G.; Mavridis, M.; Mayoral, M. L.; Mazon, D.; McCarthy, P.; McAdams, R.; McArdle, G.; McCarthy, P.; McClements, K.; McDermott, R.; McMillan, B.; Meisl, G.; Merle, A.; Meyer, O.; Milanesio, D.; Militello, F.; Miron, I. G.; Mitosinkova, K.; Mlynar, J.; Mlynek, A.; Molina, D.; Molina, P.; Monakhov, I.; Morales, J.; Moreau, D.; Morel, P.; Moret, J. M.; Moro, A.; Moulton, D.; Müller, H. W.; Nabais, F.; Nardon, E.; Naulin, V.; Nemes-Czopf, A.; Nespoli, F.; Neu, R.; Nielsen, A. H.; Nielsen, S. K.; Nikolaeva, V.; Nimb, S.; Nocente, M.; Nouailletas, R.; Nowak, S.; Oberkofler, M.; Oberparleiter, M.; Ochoukov, R.; Odstrčil, T.; Olsen, J.; Omotani, J.; O'Mullane, M. G.; Orain, F.; Osterman, N.; Paccagnella, R.; Pamela, S.; Pangione, L.; Panjan, M.; Papp, G.; Papřok, R.; Parail, V.; Parra, F. I.; Pau, A.; Pautasso, G.; Pehkonen, S. P.; Pereira, A.; Perelli Cippo, E.; Pericoli Ridolfini, V.; Peterka, M.; Petersson, P.; Petrzilka, V.; Piovesan, P.; Piron, C.; Pironti, A.; Pisano, F.; Pisokas, T.; Pitts, R.; Ploumistakis, I.; Plyusnin, V.; Pokol, G.; Poljak, D.; Pölöskei, P.; Popovic, Z.; Pór, G.; Porte, L.; Potzel, S.; Predebon, I.; Preynas, M.; Primc, G.; Pucella, G.; Puiatti, M. E.; Pütterich, T.; Rack, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Rasmussen, J.; Rattá, G. A.; Ratynskaia, S.; Ravera, G.; Réfy, D.; Reich, M.; Reimerdes, H.; Reimold, F.; Reinke, M.; Reiser, D.; Resnik, M.; Reux, C.; Ripamonti, D.; Rittich, D.; Riva, G.; Rodriguez-Ramos, M.; Rohde, V.; Rosato, J.; Ryter, F.; Saarelma, S.; Sabot, R.; Saint-Laurent, F.; Salewski, M.; Salmi, A.; Samaddar, D.; Sanchis-Sanchez, L.; Santos, J.; Sauter, O.; Scannell, R.; Scheffer, M.; Schneider, M.; Schneider, B.; Schneider, P.; Schneller, M.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Seidl, J.; Sertoli, M.; Šesnić, S.; Shabbir, A.; Shalpegin, A.; Shanahan, B.; Sharapov, S.; Sheikh, U.; Sias, G.; Sieglin, B.; Silva, C.; Silva, A.; Silva Fuglister, M.; Simpson, J.; Snicker, A.; Sommariva, C.; Sozzi, C.; Spagnolo, S.; Spizzo, G.; Spolaore, M.; Stange, T.; Stejner Pedersen, M.; Stepanov, I.; Stober, J.; Strand, P.; Šušnjara, A.; Suttrop, W.; Szepesi, T.; Tál, B.; Tala, T.; Tamain, P.; Tardini, G.; Tardocchi, M.; Teplukhina, A.; Terranova, D.; Testa, D.; Theiler, C.; Thornton, A.; Tolias, P.; Tophj, L.; Treutterer, W.; Trevisan, G. L.; Tripsky, M.; Tsironis, C.; Tsui, C.; Tudisco, O.; Uccello, A.; Urban, J.; Valisa, M.; Vallejos, P.; Valovic, M.; Van Den Brand, H.; Vanovac, B.; Varoutis, S.; Vartanian, S.; Vega, J.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vicente, J.; Viezzer, E.; Vignitchouk, L.; Vijvers, W.A.J.; Villone, F.; Viola, B.; Vlahos, L.; Voitsekhovitch, I.; Vondráček, P.; Vu, N. M.T.; Wagner, D.; Walkden, N.; Wang, N.; Wauters, T.; Weiland, M.; Weinzettl, V.; Westerhof, E.; Wiesenberger, M.; Willensdorfer, M.; Wischmeier, M.; Wodniak, I.; Wolfrum, E.; Yadykin, D.; Zagórski, R.; Zammuto, I.; Zanca, P.; Zaplotnik, R.; Zestanakis, P.; Zhang, W.; Zoletnik, S.; Zuin, M.
2017-01-01
Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine
DEFF Research Database (Denmark)
Meyer, H.; Eich, T.; Beurskens, M.
2017-01-01
Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine ...
Energy balance in the TCA tokamak plasma with Alfven wave heating
International Nuclear Information System (INIS)
Ding Ning; Qu Wenxiao; Huang Li; Long Yongxing; Qiu Xiaoming
1993-01-01
The energy balance in TCA tokamak plasma with Alfven wave heating is studied, in which the equivalent electron thermal conductivity is determined by using the profile consistency principle. The results are in good agreement with experiments. It is shown that this method is applicable to various devices and other heating methods
Simulations of edge and scrape off layer turbulence in mega ampere spherical tokamak plasmas
DEFF Research Database (Denmark)
Militello, F; Fundamenski, W; Naulin, Volker
2012-01-01
The L-mode interchange turbulence in the edge and scrape-off-layer (SOL) of the tight aspect ratio tokamak MAST is investigated numerically. The dynamics of the boundary plasma are studied using the 2D drift-fluid code ESEL, which has previously shown good agreement with large aspect ratio machin...
Effects of density asymmetries on heavy-impurity transport in a rotating tokamak-plasma
International Nuclear Information System (INIS)
Romanelli, M.; Ottaviani, M.
1997-12-01
The transport equations of heavy trace-impurities in a Tokamak plasma with strong toroidal rotation have been studied analytically in the collisional regime. It is found that the poloidal asymmetry of the impurity-density, which occurs because of the rotation, brings about a large enhancement of the diffusivity and indeed of the pinch velocity above the conventional Pfirsh-Schlueter values. (author)
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.
Limiter/vacuum system for plasma impurity control and exhaust in tokamaks
International Nuclear Information System (INIS)
Abdou, M.; Brooks, J.; Mattas, R.
1980-01-01
A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification
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
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
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)
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.)
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.
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)
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)
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
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
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)
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
Two-body similarity and its violation in tokamak edge plasmas
International Nuclear Information System (INIS)
Catto, P.J.; Knoll, D.A.; Krasheninnikov, S.I.
1996-01-01
Scaling laws found under the assumption that two-body collisions dominate can be effectively used to benchmark complex multi-dimensional codes dedicated to investigating tokamak edge plasmas. The applicability of such scaling laws to the interpretation of experimental data, however, is found to be restricted to the relatively low plasma densities ( 19 m -3 ) at which multistep processes, which break the two-body collision approximation, are unimportant. copyright 1996 American Institute of Physics
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
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)
Generación y dinámica de electrones runaway en plasmas tokamak
Fernández Gómez, Isabel
2016-01-01
La dinámica y generación de electrones runaway en plasmas tokamak constituye el tema central de esta tesis. En un tokamak, el fenómeno runaway es el resultado de la existencia de un campo eléctrico en dirección toroidal. Aquellos electrones cuya velocidad excede un cierto valor crítico se aceleran de forma continua, ya que la e ciencia de las colisiones para disipar la energía ganada en el campo disminuye con la velocidad (∼ ⁻¹) . Se tiene entonces lo que se conoce como un electrón runaway. ...
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)
Control, pressure perturbations, displacements, and disruptions in highly elongated tokamak plasmas
International Nuclear Information System (INIS)
Marcus, F.B.; Hofmann, F.; Tonetti, G.; Jardin, S.C.; Noll, P.
1989-06-01
The control and evolution of highly elongated tokamak plasmas with large growth rates are simulated with the axisymmetric, resistive MHD code TSC in the geometry of the TCV tokamak. Pressure perturbations such as sawteeth and externally programmed displacements create initial velocity perturbations which may be stabilized by low power, rapid response coils inside the passively stabilizing vacuum vessel, together with slower shaping coils outside the vessel. Vertical disruption induced voltages and forces on the rapid coils and vessel are investigated, and a model is proposed for an additional vertical force due to poloidal currents. (author) 6 figs., 1 tab., 26 refs
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
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
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)
International Nuclear Information System (INIS)
Xu Weidong; Xuan Weimin; Yao Lieying; Wang Yingqiao
2012-01-01
The high voltage power supply (HVPS) based on pulse step modulation (PSM) has already been developed for the auxiliary heating system on HL-2A. This power supply consists of many switch power supplies, and its output voltage can be obtained by modulating their delay time and pulse widths. The PSM topology and control principle are presented in this paper. The simple algorithms for the control system are explained clearly. The switch power supply (SPS) module has been built and the test results show it can meet the requirements of the auxiliary heating system. Now, 112 SPS modules and the whole system have already been developed. Its maximum output is about 72 kV/93 A. The protection time is less than 5 μs. The different outputs of this power supply are used for the electron cyclotron resonant heating (ECRH) system with different duty ratios. The experimental results of the entire system are presented. The results indicate that the whole system can meet the requirements of the auxiliary heating system on HL-2A.
Poloidal rotation induced by injecting lower hybrid waves in tokamak plasma edge
International Nuclear Information System (INIS)
Jiao Yiming; Gao Qingdi; Shi Bingren
2001-01-01
The poloidal rotation of the magnetized edge plasma in tokamak driven by the ponderomotive force which is generated by injecting lower hybrid wave (LHW) electric field has been studied. The LHW is launched from a waveguide in the plasma edge, and by Brambilla's grill theory, analytic expressions for the wave electric field in the slab model of an inhomogeneous cold plasma have been derived. It is shown that a strong wave electric field will be generated in the plasma edge by injecting LH wave of the power in MW magnitude, and this electric field will induce a poloidal rotation with a sheared poloidal velocity
Neutral molecules in tokamak edge plasma - role of vibrationally excited hydrogen molecules
International Nuclear Information System (INIS)
Cadez, I.; Cercek, M.; Pelicon, P.; Razpet, A.
2003-01-01
The role of neutral molecules in edge plasma is discussed with special emphasis on the vibrationally excited hydrogen. Neutral molecules are formed mostly by surface processes on the walls and then released to the edge plasma where they take part in volumetric reactions with other particles. Typically these molecules are formed in excited states and data are needed for their reactions on the wall and in the volume. Processes in edge plasma determine particle and energy flux what is especially critical issue in tokamak divertor region. Various cross sections and reaction rates are needed for modelling edge plasma and its interaction with walls. (author)
The Role of an Electric Field in the Formation of a Detached Regime in Tokamak Plasma
Senichenkov, I.; Kaveeva, E.; Rozhansky, V.; Sytova, E.; Veselova, I.; Voskoboynikov, S.; Coster, D.
2018-03-01
Modeling of the transition to the detachment of ASDEX Upgrade tokamak plasma with increasing density is performed using the SOLPS-ITER numerical code with a self-consistent account of drifts and currents. Their role in plasma redistribution both in the confinement region and in the scrape-off layer (SOL) is investigated. The mechanism of high field side high-density formation in the SOL in the course of detachment is suggested. In the full detachment regime, when the cold plasma region expands above the X-point and reaches closed magnetic-flux surfaces, plasma perturbation in a confined region may lead to a change in the confinement regime.
High-power heating experiment of spherical tokamaks by use of plasma merging
International Nuclear Information System (INIS)
Ueda, Yoshinobu; Ono, Yasushi
1999-01-01
High-power heating of spherical tokamaks (STs) has been investigated experimentally by use of plasma merging effect. When two STs were coaxially collided, thermal energy of a colliding ST was injected into a target ST during short reconnection time (Alfven time). Though the thermal energy increment increased with decreasing plasma q value, thermal energy loss during the following relaxation, tended to be smaller with increasing q. The produced high-β STs had hallower current profiles and weaker paramagnetic toroidal field than those of single STs. Those heating properties indicate the plasma merging to be a promising initial heating method of ST plasmas. (author)
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.
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.)
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.)
Magnetic field structure near the plasma boundary in helical systems and divertor tokamaks
International Nuclear Information System (INIS)
Nagasaki, Kazunobu; Itoh, Kimitaka
1990-02-01
Magnetic field structure of the scrape off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall is calculated. In helical systems, the connection length, L, has a logarithmic dependence on the distance from the outermost magnetic surface or that from the residual magnetic islands. The effect of axisymmetric fields on the field structure is also determined. In divertor tokamaks, the connection length also has logarithmic properties near the separatrix. Even when the perturbations, which resonate to rational surfaces near the plasma boundary, are added, logarithmic properties still remain. We compare the connection length of torsatron/helical-heliotron systems with that of divertor tokamaks. It is found that the former is shorter than the latter by one order magnitude with similar aspect ratio. (author)
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
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
Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M
2012-01-10
Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.
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.
Can tokamaks PFC survive a single event of any plasma instabilities?
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.
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
Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models
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.
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.
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
Spectroscopy and atomic physics of highly ionized Cr, Fe, and Ni for tokamak plasmas
Feldman, U.; Doschek, G. A.; Cheng, C.-C.; Bhatia, A. K.
1980-01-01
The paper considers the spectroscopy and atomic physics for some highly ionized Cr, Fe, and Ni ions produced in tokamak plasmas. Forbidden and intersystem wavelengths for Cr and Ni ions are extrapolated and interpolated using the known wavelengths for Fe lines identified in solar-flare plasmas. Tables of transition probabilities for the B I, C I, N I, O I, and F I isoelectronic sequences are presented, and collision strengths and transition probabilities for Cr, Fe, and Ni ions of the Be I sequence are given. Similarities of tokamak and solar spectra are discussed, and it is shown how the atomic data presented may be used to determine ion abundances and electron densities in low-density plasmas.
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).
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)
High kinetic energy plasma jet generation and its injection into the Globus-M spherical tokamak
International Nuclear Information System (INIS)
Voronin, A.V.; Gusev, V.K.; Petrov, Yu.V.; Sakharov, N.V.; Abramova, K.B.; Sklyarova, E.M.; Tolstyakov, S.Yu.
2005-01-01
Progress in the theoretical and experimental development of the plasma jet source and injection of hydrogen plasma and neutral gas jets into the Globus-M spherical tokamak is discussed. An experimental test bed is described for investigation of intense plasma jets that are generated by a double-stage plasma gun consisting of an intense source for neutral gas production and a conventional pulsed coaxial accelerator. A procedure for optimizing the accelerator parameters so as to achieve the maximum possible flow velocity with a limited discharge current and a reasonable length of the coaxial electrodes is presented. The calculations are compared with experiment. Plasma jet parameters, among them pressure distribution across the jet, flow velocity, plasma density, etc, were measured. Plasma jets with densities of up to 10 22 m -3 , total numbers of accelerated particles (1-5) x 10 19 , and flow velocities of 50-100 km s -1 were successfully injected into the plasma column of the Globus-M tokamak. Interferometric and Thomson scattering measurements confirmed deep jet penetration and a fast density rise ( 19 to 1 x 10 19 ) did not result in plasma degradation
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
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.)
Anomalous Convection Reversal due to Turbulence Transition in Tokamak Plasmas
International Nuclear Information System (INIS)
Sun Tian-Tian; Chen Shao-Yong; Huang Jie; Mou Mao-Lin; Tang Chang-Jian; Wang Zhan-Hui; Peng Xiao-Dong
2015-01-01
A critical physical model, based on the ion temperature gradient (ITG) mode and the trapped electron mode (TEM), trying to explain the spatio-temporal dynamics of anomalous particle convection reversal (i.e., the particle convective flux reverses from inward to outward), is developed numerically. The dependence of density peaking and profile shape on the particle convection is studied. Only the inward pinch could lead to the increase of the density peaking. The validation of the critical model is also analyzed. A comparison of the estimates calculated by the model and the experimental results from the Tore Supra tokamak shows that they are qualitatively both consistent. (paper)
Search for coherent structure within tokamak plasma turbulence
International Nuclear Information System (INIS)
Zweben, S.J.
1985-01-01
Two-dimensional tokamak edge density turbulence data are examined for possible coherent or organized structure. The spatial patterns of density fluctuations n appear to consist of localized ''blobs'' of relatively high or low density which can move irregularly both radially and poloidally through the edge region. However, a statistical analysis of the lifetime, area, direction, speed, and amplitude of these blobs does not as yet suggest any organized structure associated with the blobs beyond that which can be described by time-averaged correlation functions
Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment
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
Plasma heating and fuelling in the Globus-M spherical tokamak
International Nuclear Information System (INIS)
Gusev, V.K.; Barsukov, A.G.; Belyakov, V.A.
2005-01-01
The results of the last two years of plasma investigations at Globus-M are presented. Described are improvements helping to achieve high performance OH plasmas, which are used as the target for auxiliary heating and fuelling experiments. Increased energy content, high beta poloidal and good confinement are reported. Experiments on NBI plasma heating with a wide range of plasma parameters were performed. Some results are presented and analyzed. Experiments on RF plasma heating in the frequency range of fundamental ion cyclotron harmonics are described. In some experiments which were performed for the first time in spherical tokamaks, promising results were achieved. Noticeable ion heating was recorded at low launched power and a high concentration of hydrogen minority in deuterium plasmas. Simulations of RF wave absorption are briefly discussed. Described also are modification of the plasma gun and test-stand experiments. Fuelling experiments performed at Globus-M are discussed. (author)
Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC
Energy Technology Data Exchange (ETDEWEB)
Matsukawa, M. E-mail: matsukaw@naka.jaeri.go.jp; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T
2003-09-01
The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control.
Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC
International Nuclear Information System (INIS)
Matsukawa, M.; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T.
2003-01-01
The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control
International Nuclear Information System (INIS)
Hassanein, A.; Konkashbaev, I.
1999-01-01
Surface and structural damage to plasma-facing components (PFCs) due to the frequent loss of plasma confinement remains a serious problem for the tokamak reactor concept. The deposited plasma energy causes significant surface erosion, possible structural failure, and frequent plasma contamination. Surface damage consists of vaporization, spallation, and liquid splatter of metallic materials. Structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. To evaluate the lifetimes of plasma-facing materials and nearby components and to predict the various forms of damage that they experience, comprehensive models (contained in the HEIGHTS computer simulation package) are developed, integrated self-consistently, and enhanced. Splashing mechanisms such as bubble boiling and various liquid magnetohydrodynamic instabilities and brittle destruction mechanisms of nonmelting materials are being examined. The design requirements and implications of plasma-facing and nearby components are discussed, along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components
International Nuclear Information System (INIS)
Wang, Jian-Hua.
1990-01-01
Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (n e ∼ 1 - 5 x 10 14 (cm -3 ), T e ∼ 4 - 10 (eV), B t ∼ 0.2 - 0.4(T)). Carbon impurity light, mainly the strong lines of C II (4267A, emitted by the C + ions) and C III (4647A, emitted by the C ++ ions), is formed by the ablation or sputtering of plasma ions and by the discharge of the carbon probe itself. The diffusion transport of the carbon ions is modeled by measuring the space-and-time dependent spectral light emission of the carbon ions with a collimated optical beam and photomultiplier. The point of emission can be observed in such a way as to sample regions along and transverse to the toroidal magnetic field. The carbon ion diffusion coefficients are obtained by fitting the data to a diffusion transport model. It is found that the diffusion of the carbon ions is ''classical'' and is controlled by the high collisionality of the HBT plasma; the diffusion is a two-dimensional problem and the expected dependence on the charge of the impurity ion is observed. The measurement of the spatial distribution of the H α emissivity was obtained by inverting the light signals from a 4-channel polychromator, the data were used to calculate the minor-radial influx, the density, and the recycling time of neutral hydrogen atoms or molecules. The calculation shows that the particle recycling time τ p is comparable with the plasma energy confinement time τ E ; therefore, the recycling of the hot plasma ions with the cold neutrals from the walls is one of the main mechanisms for loss of plasma energy
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)
Impurity injection into tokamak plasmas by erosion probes
International Nuclear Information System (INIS)
Hildebrandt, D.; Bakos, J.S.; Buerger, G.; Paszti, F.; Petravich, G.
1987-08-01
Exposing special erosion probes into the edge plasma of MT-1 the impurities Li and Ti were released and contaminated the plasma. By the use of collector probes the torodial transport of these impurities were investigated. The results indicate a preferential impurity flow into codirection of the plasma current. However, the asymmetric component of this flow is much larger than expected from the toroidal drift correlated to the plasma current. (author)
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)
International Nuclear Information System (INIS)
Welander, A.S.; Deranian, R.D.; Humphreys, D.A.; Leuer, J.A.; Walker, M.L.
2005-01-01
Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII-D, the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone.In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. The linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution
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
Real-time software for the COMPASS tokamak plasma control
International Nuclear Information System (INIS)
Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Sartori, F.; Janky, F.; Cahyna, P.; Hron, M.; Panek, R.
2010-01-01
The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 μs. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.
Real-time software for the COMPASS tokamak plasma control
Energy Technology Data Exchange (ETDEWEB)
Valcarcel, D.F., E-mail: danielv@ipfn.ist.utl.p [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic)
2010-07-15
The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 {mu}s. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.
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